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C.85 - Rncm-hircm{r Nuclear
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AIRCRAFT NUCLEAR PROPULSION PROJECT

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QUARTERLY PROGRESS REPORT G\ @’”
FOR PERIOD ENDING MARCH 31, 1958

 

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OAK RIDGE NATIONAL LABORATORY
operated by
UNION CARBIDE CORPORATION

for the

CENTRAL RESEARCH LIBRARY U.S. ATOMIC ENERGY COMMISSION
DOCUMENT COLLECTION

LIBRARY LOAN COPY
DO NOT TRANSFER TO ANOTHER PERSON

If you wish someone else to see this
document, send in name with document
and the library will arrange a loan.

 
 

ORNL.-.2517
C-85 = Reactors-Aircraft Nuclear
Propulsion Systems

This document consists of 146 pages.

Copy/y‘] of 219 copies. Series A,

 

Contract No. W=7405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

For Period Ending March 31, 1958

A, J. Miller, Project Coordinator

DATE ISSUED

AUG 201958

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tonnuuo
operated

UNION CARBIDE CORPORATION
U.S. ATOMIC ENERGY COMMISSION

  
  

      

MARTIN MARIETTA ENERGY SYSTEMS LIBRAR

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FOREWORD

The ORNL-ANP program provides research and development support in shielding,
reactor materials, and reactor engineering to organizations engaged in the development
of air-cooled and liquid-metal-cooled reactors for aircraft propulsion. The work described
here is basic to or in direct support of investigations under way at General Electric
Company, Aircraft Nuclear Propulsion Department, and Pratt & Whitney Aircraft Division,
United Aircraft Corporation. This report is divided into four major parts: 1. Metallurgy,

2. Chemistry and Radiation Damage, 3. Engineering, and 4. Shielding.

 

 
 

;’ifi o ?

 

 

 
INTRODUCTION

The current ANP program at ORNL consists
primarily of supporting research and developmental
work for the General Electric Company, Aircraft
Propulsion Department, on air-cooled
and for Pratt & Whitney Aircraft on
lithium-cooled reactors, Metallurgical, chemical,
and engineering information required to develop
and evaluate materials and components for use in
specific systems is being obtained. The work
in support of lithium-cooled reactor development is
providing information on factors affecting the
stability of high-temperature columbium-lithium
systems, factors affectingthe mechanical properties
of columbium, basic hydrodynamic and heat trans-
fer characteristics of lithium systems, and com-
ponents for use in lithium systems., Work under
way on moderator materials for use at high temper-
atures, shielding materials, and the production of
low-oxygen-content yttrium metal is related to the
requirements of both types of reactor.

A comprehensive shielding program is being
carried out to aid in the development of lightweight
and crew shields for nuclear-powered

Nuclear
reactors

reactor

 

aircraft.  The major experimental facilities em-
ployed are the Lid Tank Shielding Facility (L TSF),
which provides a beam of radiation for attenuation
measurements on materials in slab geometry im-
mersed in water; the Bulk Shielding Facility
(BSF), which provides a reactor for attenuation
measurements in three-dimensional geometry of
configurations immersed in water; and the Tower
Shielding Facility (TSF), at which operating re-
actors, reactor shields, and crew shields can be
suspended as high as 200 ft in the air, Much of
the experimentation at these facilities is on
material configurations and shield mockups speci-
fied by GE or P&W in connection with their current
design problems. Measurements are also made to
obtain fundamental shielding data as part of a
basic program of study of the theory of radiation
attenuation and the development of shield design
methods.

Details of the progress made on this program
during the past three months are presented in this
report., Future progress reports will be on a semi-
annual basis.

 
 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

SUMMARY

PART 1. METALLURGY

1.1. Fabrication

Effort is being devoted to determining the effects
of impurities, such as oxygen, nitrogen, hydrogen,
and carbon, on the fabricability, weldability,
mechanical properties, and corrosion resistance
of columbium. In comparisons of cold-worked
powder-metallurgy and arc-cast material, the form
in"which the impurities are present was found to be
as important as the quantity of the impurities,
Studies of the effects of various heat treatments
initiated, and the
preliminary results indicated that nitrogen is the
impurity most influential in heat treatment effects.

Work was continued on the fabrication of co-
lumbium tube shells. Three arc-cast billets were
extruded satisfactorily and are to be drawn into
tubing. Additional billets are being obtained for
determining optimum extrusion temperatures and
cladding procedures,

Coatings of brazing alloys applied by flame
spraying are being tested for oxidation protection
of columbium. Both Coast Metals brazing alloys
52 and 53 were found to provide satisfactory
protection in 1000-hr oxidation tests at 1700°F,
Subsequent metallographic examination, however,
revealed cracks in the coatings that could have
led to rapid failure. Studies of the reaction rates
between low-pressure gases and columbium were
initiated

on impure columbium were

in order to determine the permissible
limits of gaseous contaminants in the environment
when columbium is exposed at high temperatures,

Developmental work on methods and apparatus
for the production of low-oxygen-content yttrium
metal was continued. The method now being
studied involves the lithium reduction of a low-
melting mixture of the fluorides of yttrium, mag-
nesium, and lithium to produce an yttrium-magnesium
alloy from which yttrium sponge is produced by
vacuum distillation. Castings prepared from the
sponge by arc melting under an argon atmosphere
with a tungsten electrode were analyzed for im-
purity content, The chemical analyses indicated
oxygen contfents af 600 to 3000 ppm and nitrogen
contents of 50 to 17,000 ppm. Experiments are

 

under way to determine the relationship between
impurity content and tensile properties.

An electron-beam method of consolidating the
yttrium-magnesium alloy directly into a finished
billet is being investigated. This method, if
successful, could eliminate the vacuum-distillation
and arc-melting steps. Also, equipment is being
assembled for hydriding yttrium,

1.2. Corrosion

Methods are being studied for purifying the
lithium required in projects in support of both the
air-cycle and the lithium-cooled reactors, Lithium
as pure as it is practical to produce is needed for
the reduction step in the production of low-oxygen-
content yttrium metal, as well as for studies of
the effects of minor amounts of oxide and nitride
on the corrosion of reactor materials, Determi-
nations of the solubility of nitrogen and oxygen in
lithium and studies of various purification tech-
niques indicated that gettering was a more satis-
factory purification technique than low-temperature
filtration or vacuum distillation, Initial gettering
tests with zirconium, titanium, and a mixture of
titanium and calcium indicated significant re-
ductions of the nitrogen content of lithium, but
these gettering materials had little effect on the
oxygen content,  Further experiments will be
performed with other metals which have oxides
whose free-energy values indicate that they are
suitable as getters for oxygen. The extensive
oxidation of an yttrium metal specimen upon ex-

posure to lithium suggested the possibility of
using yttrium metal in the purification of the
lithium required in the reduction step of the

yttrium production process.

Two zirconium-base alloys being investigated for
use as weld-filler rod in the fabrication of ductile
saddle welds in columbium tubing were tested for
corrosion resistance in lithium, Static and thermal-
convection loop tests showed weld joints made
with both an 85% Zr—-15% Cb and an 80% Zr-15%
Cb-5% Mo alloy to have excellent corrosion
resistance.

Thermal-convection loop tests of lithium-
columbium systems have thus far been of limited

vii

 
, Y

duration because of the high rate of penetration
of the longitudinal weld in the wall of the columbium
tubing by the lithium metal, Two loops of sintered
columbium tubing with inserts of arc-cast material
showed very little attack away from the longitudinal
weld, but there was severe attack in the heat-
affected and weld zones.

Static tests of lithium-filled columbium tubes
were made in order to determine whether the en-
vironment in which the thermal-convection loop
tests were made had an effect on the weld attack.
Four capsules, two in an argon atmosphere and
the other two in a dynamic vacuum of less than
5 u, were attacked similarly. The weld and heat-
affected zones of all four capsules were heavily
attacked, and there were several cases of complete
The base material away from the
weld zone was unattacked.

Experimental work on a lead-lithium alloy con-
taining 0.69 wt % lithium - a potential shielding
material — was completed. An experimental
shield was prepared for studies at the Bulk Shield-
ing Facility.

penetration,

1.3. Welding and Brazing

The effect of oxidation on brazed joints is being
investigated. A nickel-rich solid-solution layer
forms on the fillet surfaces of joints brazed with
the Coast Metals nickel-base brazing alloys and
the depleted region is quite ductile. Thus it
appears that the ductility of oxidized joints of
this type should improve during service. Studies
of the rate of depletion of microcanstituents during
high-temperature oxidation are under way.

Refractory-metal brdzing alloys for use in
lithium-cooled reactor applications are being in-
vestigated, since the nickel-base brazing alloys
have poor resistance to lithium, A survey was
made of the columbium and zirconium binary
systems of potential interest, and other systems
will be studied.

1.4. Mechanical Properties

Investigations were continued of the high-
temperature mechanical properties of structural
The testing programs include mechanical
and thermal strain cycling, mechanical stress
cycling (fatigue), creep in bending, and creep
under multiaxial stresses. Inconel was chosen

as the first test material because its tensile,

metals.

uniaxial creep, and relaxation characteristics are

vili

e 37

well known and could be used in the analysis of
its behavior under more complex stress situations.

In mechanical strain cycling, emphasis is placed
on evaluating the influence of frequency of cycling
on the number of cycles to failure. The extent to
which the cycle time influences the number of
cycles to rupture is dependent on the plastic
strain per cycle and the temperature. A com-
parison of the number of cycles to rupture at
low strains for cycle times of 2 and 30 min re-
vealed that long cycle times decrease the strain
absorption capacity. This effect becomes more
apparent at the low strain-per-cycle values ‘and is
also magnified by increasing temperature, A
severe frequency effect, for example, was observed
at 1% strain and at 1500°F; fine-grained material
endured approximately 700 cycles at 2 min per
cycle and only 160 cycles at 30 min per cycle.
At 1300°F, however, 2- and 30-min cycles yielded
the same number of cycles to rupture at 1% strain,

A series of thermal strain cycling tests at a
mean temperature of 1500°F were completed at
the University of Alabama. The correlation between
mechanical and thermal cycling data was very
Lack of agreement between the tests is
attributed to the unreliability of the techniques
presently used to measure the plastic strain per
cycle in the thermal cycling tests,

Fatigue studies were conducted by Battelle
Memorial Institute, Data obtained at 1600°F on
specimens stress cycled at 1 and 10 cps revealed
a similar frequency effect to that observed in the
strain cycling tests conducted at ORNL, Specimens
cycled at the high frequency endured 10 times
more cycles before failure than did the specimens
in comparable lower frequency tests.

The effect of cyclic stresses on creep properties
was also investigated. Small alternating stresses
were beneficial to the creep properties. The time
to rupture showed a slight increase, and the creep
rate decreased in comparison with pure creep
properties, Upon further increasing the alternating
stress, the time to rupture decreased. Conventional
Goodman-type diagrams, which relate time to
rupture to the stress ratio, were obtained for the
two frequencies, 1 and 10 cps.

Results from creep-bending tests at 1500°F
the hypothesis that creep-bending be-
havior can be predicted from tensile creep data,
As postulated, creep in the outer beam fibers
will follow curves in tensile creep tests at the

poor,

support

 

 
stress initially present in the outer fibers. How-
ever, after the first few hours, the bending creep
rate correlates more closely with tensile creep
data corresponding to a lower stress. This indi-
cates that the bending stresses in the beam are
redistributed through relaxation and that the re-
distribution results in a lower stress in the outer
fiber of the beams.

Multiaxially stressed tubular specimens were
creep tested at 1500°F. |t was found that the creep
rate in the axial direction was not influenced by
tangential stress in cases where the axial-to-
tangential stress ratio, O’Z/Ort, was greater than
0.5. In other words, the axial stress alone de-
termined the axial creep rate. At the stress ratio
of 0.5, no axial creep was observed, although a
tensile stress was present in the axial direction,
At stress ratios, 0 /o, from 0.5 to -1.0, com-
pressive axial creep was observed. Incompression,
however, creep rates were accelerated by tan-

gential stress.

In contrast to the creep rates, time to rupture was
greatly affected by tangential stress. As the
stress ratio decreased from « to 1.0, for example,
when one of the principal stresses was held at
6000 psi, the time to rupture decreased from 350 to
90 hr. At ratios below 1.0, where only the axial
stress was varied, the variance of time to rupture
with stress ratio was only slight, although the time
to rupture continued to decrease as the stress
ratio decreased,
occurred at the stress ratio —1.0,

The minimum time to rupture

At present neither the creep nor rupture behavior
can be related to the common mechanical theories
of flow and fracture, However, a number of tenta-
tive conclusions can be drawn from the data, It
seems likely that deformation rates under multi-
axial stresses are controlled by the maximum
normal stress in the axial direction in the range
of stress ratios from 0.5 to @, This implies that
data from conventional uniaxial creep tests are
valid bases for design under these conditions.
However, the elongation to which a material may
be stretched before failure is markedly affected by
changes in stress distribution, and it decreases
greatly with increased tangential stress. Recog-
nition of this fact by the designer may alter the
selection of a limiting strain.

 

1.5. Ceramics

Dense beryllium oxide is being studied for
use as a neutron reflector material in air-cycle
The possibility of adding boron as
a suppressor for secondary gamma rays is also
being investigated. Specimens containing ZrB,,
TaB,, TiB,, GrB,, HfB,, CeBG, and BN were
prepared, and the specimens containing ZrB, and
Hf82 were found to be much superior to the other
combinations with respect to oxidation resistance
at 1300°C. Additives are being sought that will
densify extruded and cold-pressed BeO bodies and
possibly, but not necessarily, fulfill the boron con-
tent requirements,

reactors,

1.6. Nondestructive Testing

The study of eddy-current techniques for measur-
ing cladding thicknesses was continued. Corre-
lation of the measurements with the results of
metallographic examinations is under way. Also,
the possibility of measuring the thickness of a
nonmagnetic cladding material on a magnetic core
is being investigated. Measurements of aluminum
coating thicknesses on steel in the absence of an
intermetallic layer can be made with an accuracy
of 0,0001 in. or greater,

During investigation of the problem of measuring
the thickness of aluminum coatings on steel, a
new technique was developed which shows promise
of eliminating the inaccuracies introduced by [ift-
off or poor coupling between the probe coil and the
part. A method that is independent of |ift-off
variations will be adaptable to automatic scanning.

An ultrasonic phenomenon Lamb
waves, which are basically a characteristic elastic
vibration established in a thin metal plate, is being
investigated for possible use in the detection of
laminations, The Lamb waves are generated in
thin metal sections by directing a beam of ultra-
sonic energy at particular incident angles to the
sheet surface.

known as

Equipment has been obtained and is being tested
for measurements of ultrasonic attenuation and
velocity. Changes in ultrasonic attenuvation as a
function of changes in metal structures will be
examined, and methods for valid vultrasonic in-
spection of welded structures will be developed.

 
PART 2. CHEMISTRY AND RADIATION DAMAGE

2.1. Materials Chemistry

The methods developed for the preparation of
the LiF-MgF ,-YF, mixture required in the pro-
duction of yttrium metal were analyzed in terms
of possible large-scale production in available
equipment. Since scale-up of the process appears
to present no particular difficulties, a pilot plant
in which the available equipment will be utilized
was designed and is being constructed. The
pilot plant will be capable of supplying a quantity
of fluoride mixture sufficient for the production
monthly of 200 Ib of low-oxygen-content yttrium
metal.

Experimental initiated of a
possible method for the extraction of impurities

studies were

from lithium metal with molten salts. In order to
avoid contamination of the lithium by extraneous
metal, the eutectic mixture LiF-LiBr (12-88
mole %) was chosen as the extractant for preliminary
studies. In an additional step, bismuth wasused
to extract the residual lithium from the salt mixture
following the initial impurity extraction. Further
experiments mustbe performed before the effective-
ness of the method can be determined.

2.2. Analytical Chemistry

The precision of a method which has been pro-
posed for the determination of lithium oxide in
metallic lithium was evaluated by analyzing
samples of a dispersion of metallic lithium in
paraffin.  In this method the lithium sample is
dispersed in paraffin so that aliquots for analysis
can be obtained by merely cutting and weighing.
In order to determine the uniformity of the sample
dispersion in the paraffin, dispersions were pre-
pared with a Waring Blendor fitted with a stainless
steel mixing cup in which a helium atmosphere
was maintained. Samples of the solidified ma-
terial, which contained approximately 1 g of the
dispersed metal, were analyzed, The average
value for five samples was 230 ppm, with a co-
efficient of variation of 5%. These results indicate
the validity of the assumption that lithium oxide is
uniformly distributed in the lithium-paraffin dis-
persion, |

The apparatus previously used for determination
of oxides in fused mixtures of fluoride salts is
being adapted for use with BrF.SbF . This flux
should be applicable to the determination of oxides
in fused mixtures of lithium fluoride and magnesium
fluoride.

2.3. Radiation Damage

The tube-burst stress-rupture apparatus for tests
under irradiation in the MTR was completed, and
final preirradiation tests were initiated. The
apparatus differs from that used in previous MTR
irradiations in that the Inconel specimens will be
irradiated in air rather than in helium to avoid
changes in thermocouple calibration, and the
specimens will be cooled internally in order to
avoid overheating. While being irradiated the
specimens will be at 1500°F and at stresses of
3000 to 6000 psi. It is planned to continue tests
of Inconel specimens wuntil all experimental
difficulties have been resolved and basic data on
the material have been obtained.

Metallographic examination of an Inconel stress-
rupture specimen irradiated in the LITR for 600 hr
while at a stress of 2000 psi and a temperature of
1500°F showed that rupture had occurred outside
the gage length where the temperature was not
monitored. The specimen was exposed to a
fluoride fuel mixture on the inside and helium on
the outside. No evidence of increased corrosion
as a result of irradiation was found in the gage
length where the temperature was monitored, A
holding fixture for creep experiments in the ORR
pool has been installed, and a flux monitoring
facility was completed.

Irradiation~induced changes in semiconductor
diodes were studied in the light of the two princi-
pal barrier theories: the Shockley diffusion theory
and the charge recombination-generation theory,
Experimental evidence indicates that irradiation
causes a change in the predominant mechanism of
rectification of germanium p-n junctions from the
diffusion mechanism to the recombination-generation
mechanism, This shift in mechanism explains the
difficulty heretofore encountered in explaining
the radiation-induced changes in current-voltage
characteristics.

Work continued on the preparation of apparatus
for thermal-stress- and thermal-shock-resistance
testing of moderator materials at high temperatures
in the ETR. The specimens that will be irradiated
include beryllium oxide, beryllium, and yttrium
hydride.  Preirradiation tests of the specimens
will be made to determine elastic properties and
dimensions and the effect of heat cycling. The
yttrium hydride specimens will, in addition, be
examined by x-ray diffraction in order to determine
the crystal type and lattice spacing. These studies
will be repeated after irradiation to detect changes.

 

 
 

PART 3. ENGINEERING

3.1. Component Development and Testing

loop was designed for
studying the corrosion of columbium by lithium

A forced-circulation

metal circulating at high temperatures in a system
with a temperature gradient. In order to protect
the columbium tubing from air oxidation, the
auxiliary portions of the loop, that is, the fill
and drain lines, the line to the lithium expansion
tank, and the loop which includes the rotating-
magnet pump, will be clad with type 446 stainless
steel. The tubing that will be at high temperatures
in the test section will be surrounded with static
sodium which will transfer heat to or remove heat
from the columbium tubing in the hot and coid
legs of the loop. The cladding of the external
portion of the columbium loop will be joined with
the stainless steel expansion tank of the sodium
system. An all-stainless-steel loop of this design
is being constructed in which NaK will be circu-
lated in order to obtain operating data with which
to validate assumptions made in heat transfer and
flow calculations.

The centrifugal pump rotary assembly in oper-
ation under gamma irradiation in the canal at the
MTR had received an accumulated gamma-ray dose
of 1,6 x 107 rep by the end of the quarter, that is,
about one-half the planned dose. The equipment
continues to operate satisfactorily, and no gross
changes have been observed in the oil {Gulfcrest-
34) which is used to lubricate the bearing and
seal, Oil-lubricated pumps of this type are being
tested for applicability to liquid-metal circulation
in radiation fields,

Modifications were made to the primary face
seals of centrifugal pumps being tested for high-
service. Use of a Fulton-Sylphon
seal prevented accumulation of undesirable de-
posits on the gas side of the seal. A split purge
arrangement stopped the rise of vapor from the
pumped fluid and likewise prevented accumulation
of deposits on the seal face.

One phase of the development and testing of
shut-off valves for high-temperature service was
completed. A valve was developed which is
satisfactory for operation in NaK at temperatures
vp to 1300°F, Cermet combinations known to be
compatible with lithium were used as seat and
The valve will be satisfactory if
fabricated from columbium or another material

temperature

plug materials.

compatible with lithium.

3.2. Heat Transfer Studies

Experimental studies of the effect of thermal
stress cycling on structural materials were con-
tinued with the use of the pulse-pump system.
The effects of exposure time, that is, total number
of thermal cycles, on the Inconel test specimens
are being investigated. For correlation with these
studies, measurements are being made of the
attenuation of sinusoidal temperature oscillations
at the interior surface of a thick-walled Inconel
cylinder,

The study of the effect of screens in a diverging
annular channel on the thermal structure of a
fluid with internal heat generation was completed.
The wuse of variable porosity screens reduced
the amplitude of the surface temperature fluctu-
ations to less than 10% of the axial rise at both
the inner and outer channel walls. The screens
also effectively reduced flow asymmetries caused
by stoppage of one of the two pumps in the flow
system,

Initial studies of heat transfer in a liquid metal
with internal heat generation were completed.
Preliminary analyses of the data indicate that the
surface-to-fluid mean temperature differences are
greater than predicted. It thus appears that the
theory-vs-experiment discrepancies observed in
liquid-metal wall-heat-transfer systems — that is,
externally heated systems — also exist in the in-
ternally heated systems,

3.3. Instruments and Controls

A data acquisition, or digital recording, system
was obtained for use in recording and correlating
the process variables for which measuring instru-
ments will be required in a reactor system. The
information obtained with this system will be use-
ful in evaluations of the need for instruments and
the preparation of specifications for instruments.

Life tests of ORNL-designed liquid-metal-tevel
transducers were continued., Two units have
accumulated 5408 hr of operation at 1200°F in
NaK and two have accumulated 775 hr. Eight
units were tested to investigate wetting in NaK at
various temperatures. The data are being analyzed.
Two l-in. turbine flowmeters that had operated in
NaK for 2000 hr at temperatures up to 1560°F were
removed from the test loop and are being dis-
assembled for examination. The design principles
embodied in these instruments are suitable for the

 

 
design of instruments for use with high-temperature
lithium systems.

Strain gages for static measurements at temper-
atures up to 2000°F are being studied, Since
past experience with strain gages has been almost
entirely at low temperatures, a test facility for
evaluating various types of strain gages at high
temperatures is being devised,

A design of an in-pile loop in which to study the
effects of high-level radiation on thermocouples
was prepared. The test environment will be
sodium,

Tests are under way on pneumatic pressure
transmitters., Overrange pressures are being used
to determine the aging characteristics of the
instruments with respect to accuracy and over-all
safety.

PART 4. SHIELDING

4.1. Shielding Theory

In order to predict and check the results of
shielding experiments, a computation was made
of the energy and angular distribution of the gamma
rays at the surface of the BSF reactor. The calcu-
lated gamma-ray energy spectrum normal to the
surface of the reactor checked with that obtained
in previous calculations, and the first detailed
calculation of angular distribution was completed.

4,2. Lid Tank Shielding Facility

The experiment designed by GE-ANP for studying
the production of secondary gamma rays in con-
figurations containing advanced shielding ma-
terials is being continued. The latest tests were
performed to determine the optimum placement
of boral within configurations containing stainless
steel, The most effective way of reducing the
capture gamma-ray dose rates beyond the con-
figurations was to distribute the boral evenly
throughout the stainless steel and, in addition,

to place some boral behind the stainless steel.

4.3. Bulk Shielding Facility

The design of a stainless-steel-clad, U02-
fueled reactor for the BSF was completed, and a
report for submission to the Advisory Committee on
Reactor Safeguards was prepared. Mechanical tests
and analog computations carried out on the control
system indicate that it is capable of safely re-
versing positive periods as short as 8 msec.

This would be equivalent to an excess reactivity
of 1.2%; however, it is planned to limit the stain-
less steel reactivity to 0.75% until a planned
series of tests on the BSR-ll and its control
system can be performed at the SPERT Facility.

The study of the spectra of gamma radiation
associated with the fission of U237 is continuing.
It has included a second BSF determination of
prompt radiation, the results of which are in general
agreement with the previous results, although
they differ by as much as a factor of 2 in some
energy regions. Preliminary results were also
obtained in measurements of the fission-product
gamma radiation emitted during the period between
5 x 108 and 10~¢ sec after fission and during the
period between 1 and 1500 sec after fission.

A ?’/4-in.-diu by 2-in.-deep well was drilled into
the truncated end of the 9-in..dia crystal of
Nal(T1) — which will be used as a total absorption
gamma-ray spectrometer at the BSF — in order to
reduce the effects on the spectral response of the
Compton scattering near the surface where the
gamma radiation enters the crystal. In an attempt
to improve the percentage resolution, seven 3-in.-
dia photomultiplier tubes were mounted on the
crystal, In spite of this improvement, the total
spectrometer does not yet compete
favorably with other spectrometers in terms of
resolution, although its efficiency is two or three
orders of magnitude greater.

A study of the heat production in a large slab
mockup of a typical aircraft shield was undertaken

absorption

at the BSF. The mockup being investigated con-
sists of a 4-in,-thick slab of beryllium, followed by
a gamma-ray shield and 16 in. of lithium hydride.
The gamma-ray material consists of 3 in. of lead,
3 in. of iron, or 2 in. of Mallory 1000. The entire
shield is submerged in an oil-filled aluminum tank
positioned against the BSR. Measurements are
being made of the heating, dose rates, and fluxes

within the shield.

4.4. Tower Shielding Facility

An experiment in cooperation with Convair, Fort
Worth, is now being conducted at the Tower
Shielding Facility to obtain neutron and gamma-
ray dose rates and thermal-neutron fluxes under
conditions similar to those encountered with the
Nuclear Test Airplane Program but in the absence
of the airplane structure. The experiment will

also include dose rates and flux measurements

 
taken at various positions around the ASTR while
the reactor is 12,5 ft above the ground as in the
Convair ground test measurements. In order to
sort out the ground scattering effects, similar

 

measurements will be made while the reactor is |
suspended at an altitude of approximately 190 ft.
Other measurements will be made to determine
the spectra of neutrons emitted from the ASTR.

xiii

 
 

 

 

 
CONTENTS

INTRODUGCTION ...ttt st ettt et ee e e e e e es e oo

SUMMARY L.t e

.............................................................................

PART 1. METALLURGY

1.1, FABRICATION ..o

Columbium Investigations...........ccccoveviievceieiien e
Fabrication Studies ...,

Yttrium Metal Production ......cccoooiviiiiiiiiiiiie
A0y Production ...ttt ettt eneea
Evaluation of Material ...
Yttrium Hydriding ..o

1.2, CORROSION ...ttt e,

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

Lithium Purification: Methods of Reducing the Nitrogen and Oxygen Content .........c.ooeovevvennnnn..

Solubility of Nitrogen and Oxygen in Lithium....
Filtration ..o

Static Lithium Corrosion of Zirconium-Base Alloys

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

Thermal-Convection Loop Tests of Lithium in Columbium woooooeioeeeoe e

Static Lithium Corrosion of Columbium .ooveeveiveinil

1.3. WELDING AND BRAZING .......cccociiriiiiiiieiienn,

............................................................................

............................................................................

............................................................................

Oxidation of Coast Metals Brazing Alloys 52 and 53 ..o,

Development of Brazing Alloys for Lithium Service

1.4. MECHANICAL PROPERTIES ...
Strain-Cycling Investigations.....c..cccveceivinieiciniecnen.
Creep Under Bending Stresses..........cocveeeiivinceeennnn.

Creep Under Multiaxial Stresses .........cccoovivineiinnnnn.

1.5. CERAMICS ..o,
Boron Compounds ......cooveiiiiviieiiiiieees e
Boron Additions to BeO ...occoooiviiiiiiciii
Densification of BeO ...

1.6. NONDESTRUCTIVE TESTING ......ccoooviieee,
Eddy-Current Thickness Measurements ....................
Lamb Ultrasonic-Wave Studies ..o

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

............................................................................

Xv

 

 
2.1

2.2.

2.3.

3.1.

3.2,

3.3.

XVi

PART 2. CHEMISTRY AND RADIATION DAMAGE

MAT ERIALS CHEMIST RY oot ettt veeeeeerveassareesssesesrssseeatesessbsss s bbb eat e b s e an s b e sa e R e s e sat st et eatean et st sas s 53
Preparation of Charge Material for Reduction 10 YHIIUM ..cocriiciiii 53
Conversion of Y, 0,5 10 Y 4 oo 53
Preparation of MgF o oot 53
Preparation of the LiF-MgF ,-YF 3 Charge Mixture ..o, 54
The System LiF-YF 5 o 54
Extraction of Lithium Metal Impurities with Molten Salts ... 55
ANALYTICAL CHEMISTRY Lottt ot 57
Determination of Oxygen in Metallic Lithium ... 57
Determination of Oxygen in Fluoride Salts ... 57
RADIATION DAMAGE ...ttt ettt eb e eb s s b e saes e s encr e e e s e e st eas s abe e b r e s e et abeant e ste s 58
Creep and Stress Rupture Tests Under lrradiation ... 58
MT R EXPEFIMENTS .oiteiie et sr st s s e bbb bt 58
L TR EXPEIIMENES .oveiiiiiei ettt e n bbb b 59
DR R T @SS vnviereiereeeeeeeeee e eeeeaeassseaasesnsesaesaeseesssassntasstas st e abe e as s eme e saes s s e e e s s b s e e e R e e e e e A e e Renarr st et e et s 60
Radiation Effects in Electronic Components .......coevuiieeiciccnimiitieiieei it e e o 60
Irradiation Tests of Moderator Materials for Use at High Temperatures .......ccocovvvnivnniicnnn 62

PART 3. ENGINEERING

COMPONENT DEVELOPMENT AND TESTING ..o cittsse sttt s vmne e 67
Loop for Investigating Corrosion in Lithium-Columbium Systems ..., 67
PUmpP DevelOpPmMEnt ...t s 67
Irradiation Test of Bearings and Seals ..o 67
Sl [MPrOVEMENTS ..oeieieececc ettt ettt e bbb s eSS 67
CAVIEAHON TESES ooiieieeiiieetietie it ii e et e veeeeeetees s eserteeabebserse et e s a e e aean s s eheeas e e ns s enees e e ateesbesasabbeab e e s b et r et 69
Valves for Use at High Temperatures .........co.ooviiiiiiiic e 69
Thermal Stability Tests of Metal Shells ..o 69
HEAT TRANSFER STUDIES ..ottt sttt s se s st ssa s e sn e sae b e san et asna s et st sonssnsne e 72
Studies of the Effect of Thermal-Stress Cycling on Structural Materials ... 72
PUlSE-PUMP SYSTEM ..oiiiieiiiiie it s 72
Thermal Entrance Studies with Water ... e 74
Thermal Structure of Fluids Flowing in Volume-Heated Systems ..o 77
Diverging Annular Channel with Screens ... 77
Liquid-Metal EXPeriment . .....cccoccicoiiiiiiiiiiiiiiiniice et 79
INSTRUMENTATION AND CONTROLS .ottt b s s 81
Data AcQUISTTION SYSTEM .o.eiviiiieieieere ettt e b b bbb 0 81
Liquid-Metal-Level Transducers ... 81
Strain Gages for Use at High Temperatures ... 81
Test Set-Up for Study of Effect of Radiation on Thermocouples ..o 82
Pressure Transmitters for Use at High Temperatures ... 82

 

 
4.1.

4.2.

4.3.

4.4,

PART 4. SHIELDING

SHIELDING THEORY ..ottt et s e st st et s ee e ee

A Calculation of Energy Spectrum and Angular Distribution of Gamma Rays at
Surface of BSF Re@Ctor ...ttt e ettt re e

G BOMEIIY ...t e et e e et e e e e e e aneeeeanneeteteeebnneeeeaeeteeenntearnabeaeibeterbesatraranan
Thermal-Neutron FIuX .. ...t e bae e s b e ra e e smenreesbeeene o
GaMMA-RAY SOUPCES oottt et e et e e eba e s b aessan e be e bbaaearensane s aense o
Absorption Coefficients ...t e st e et e re e
BUIlUD FaCtOrS ..ot ettt e sttt ab b e sae e be s esbee s beeseeanaesrenn on
Flux ComPutGEIOn ..cooeeiieeeeeeeeeee ettt eee et e et eb e b eab et e st e besateasbaaassaasasbeensensnteensesnnnsnanas
Angular Distribution ...ttt
CONCIUSTONS .ottt ettt et e st e e te e e aeete e s st e et e sebeesaesaseanseanseentesasennasasnesnn oo
LID TANK SHIELDING FACILITY oottt sae e e
Study of Advanced Shielding Materials (GE Series) ......cococoiviiinciie e
BULK SHIELDING FACILITY ettt et ettt ekt eb ettt ebt b saeseeebene o
The BSF Stainless Steel-UO, Reactor (B O R -} e ettt
Spectra of Gamma Radiation Associated with the Fission of U237 . e,
Fission-Product Gamma Rays Emitted More Than 1 sec After Fission ...
Fission-Product Gamma Rays Emitted Between 5 x 1078 and 10~ ¢ sec
AL Er S SI0M o ieitiiiiiteieeet et ettt e s et e e ce e et oot e e etsa e e e aaeee et aeaeemseeeenteeeaenneeeeaneeeeansee nraeeeebaeeeannbenaee e
Prompt Gamma Rays ..ot ettt e et e et bb e b e e st et bt e b nn e esnen e r et s
Total Absorption Gamma-Ray SpectroSCopy ......cooiviireiiiiiiicecrie e
A Study of Radiation Heating in Shields ...........cooiiiiiiiiiiiiic e
TOWER SHIELDING FACILITY ettt ettt ettt s b et s see e n s
Aircraft Shield Test Reactor Experiment at the Tower Shielding Facility ...

xvii

 
Part 1
METALLURGY

W. D. Manly
Metallurgy Division

T. Hikido
U.S. Air Force

 
1.1. FABRICATION

J. H. Coobs
H. Inouye

D. O. Hobson
W. J. Werner

Metallurgy Division
T. Hikido, U.S. Air Force

COLUMBIUM INVESTIGATIONS

H. Inouye D. O. Hobson

Fabrication Studies

Cold Working. —~ Attempts have been made to
fabricate columbium at room temperature because
of its reactivity with gases at high temperatures.
Metal prepared by the powder metallurgy process,
in which relatively large quantities of oxides,
nitrides, and carbides were present, was suc-
cessfully fabricated, but arcecast material, in
which the impurity content was lower, was brittle
when fabricated at room temperature, It appeared
therefore that the form in which the impurities
were present in the metal was as important as
the quantity present. The impurities are taken
into solution in the melting step, and, if they
do not precipitate significantly during cooling,
a subsequent, suitable heat treatment of the
material should improve its fabricability. Thus
heat treatment of the metal at temperatures above
the solubility limit of the impurities should have
no effect on the fabricability, whereas, at a lower
temperature, precipitation should occur and result
in a ductile metal because of overaging. In order
to test this hypothesis, sections of an arc-cast
ingot prepared by the electron-beam melting process
were heat treated at 750, 850, and 1000°C for
24 hr. The lower temperatures were selected to
cause precipitation, whereas the higher temperature
was selected to be above that at which the
impurities would be retained in solution. The
results of cold rolling and bend tests of the heat-
treated specimens are presented in Table 1.1.1.
Samples from the as-cast ingot and the sample
annealed at 1000°C could not be rolled, but the
other samples were cold rolled to foil. On the
basis of the solubility data for oxygen! and
nitrogen? in columbium, it appears that the element
which causes the observed heat-treatment effects

 

VA. U. Seybolt, J. Metals 6, 77476 (1954)
2C. Y. Ang and C. Wert, J. Metals 5, 1032—36 (1953).

is nitrogen. The various specimens were identical,
metallographically, after heat treatment.
Extrusion. — Work has continued on the fab-
rication of columbium tube shells. Three arc-cast
billets were extruded satisfactorily and will be
shipped to Superior Tube Company to be drawn
into tubing. The three billets, all from Superior
Tube Company, were copper plated for protection,
heated in argon, and extruded at a 7:1 ratio.
The first billet was extruded at 900°C, and,
presumably as a result of contamination caused
by pinholes in the copper plating, the extrusion
had numerous small, deep cracks across the
surface. The second billet was extruded at
815°C. The surface was fairly good, but there
was a little roughness as a result of large grains
and of some glass lubricant having been carried
through the die unmelted. The extrusion pressure
was approximately 60 tsi, that is, well within
the limits of the press. The third billet was
extruded at 980°C without the glass lubricant,
and an extrusion pressure of only 53 tsi was

required. The surface condition was very good.

Toble 1.1.1, Effect of Heat Treotment on the
Fabricability of Cast Columbium?*

Heat treatment: 24 hr in vacuum of 10> mm Hg

at indicated temperature

 

Bend Angle

Heat Treatment

 

Cold-Rolling
Temperature at Fracture** Results
©c) (deg}
As cast 127 Brittle fracture an
first pass
750 180 Rolled to foil
850 180 Rolled to foil
1000 Brittle fracture on
first pass

 

*Impurity analysis: C, 0,01%; 02, 0.12%; H,, 0.001%;
N,, 0.032%.

**Specimen dimensions: 0.125 x 0.340 < 1 in.
ANP PROJECT PROGRESS REPORT

Twelve columbium billets have been ordered
for future work. These will be used to determine
the optimum extrusion temperatures and the best
cladding procedures. It may be possible by rapid
heating in an induction furnace to extrude bare
billets.

Oxidation Protection

Some work has continued on the oxidation
protection of columbium.  Satisfactory results
were obtained in oxidation tests for 1000-hr

periods at 1700°F of specimens coated with
Coast Metals alloys 52 and 53. The specimens
were coated by flame-spraying with the powdered
alloy and were then heated in a purified helium
atmosphere to a temperature just below the brazing
temperature,

A specimen coated with alloy 53 showed a
weight increase of 1.1 mg/cm? in 200 hr, after
which the weight decreased slowly to a total
weight gain of 0.6 mg/cm? after 1000 hr. This
weight change compares quite favorably with
that which would occur in the same period if
the specimen were type 316 stainless steel.
At 1600°F, a type 316 stainless steel specimen
would gain about 1.5 mg/em? in 1000 hr.

A specimen coated with alloy 52 showed a
weight increase during the first 24 hr and then
lost weight for 100 hr, at which time the total
weight change was a loss of 1.2 mg/cm2. After
300 hr the specimen began to gain weight again
and it was still gaining after 1000 hr.

Metallographic studies of the coatings showed
a large amount of surface oxidation, some reaction
loyer formation, and a few diffusion voids at the
interface. The Coast Metals coatings were cracked
in several places, but hardness readings showed
no hardness increases in the columbium near the
cracks. The specimen coated with alloy 53 had
an average hardness of 110 DPH, while the speci-
men coated with alloy 52 had a hardness of 78.7
DPH. The coatings ranged in hardness from 250
to 750 DPH.

Oxidation tests were also run on two columbium
tubes coated with alloy 53 and filled with lithium.
The specimens were tested at 1500°F, and they
failed because of incomplete protection by the
The heat treatment evidently did
not cause complete flow of the braze metal.

brazed coating.

Reaction Rates

The reaction rates between low-pressure gases
and columbium are being studied to determine the
permissible limits of gaseous contaminants when
columbium is exposed at high temperatures. An
apparatus is being constructed that is designed for
determining reaction rates of columbium with the
pure gases at pressure levels of 0.01 to 1000 z Hg.
In addition to providing a means for establishing
the required purity of the atmosphere during high-
temperature testing, the apparatus will be used
for the preparation of samples for the determination
of the effect of a specific impurity on the proper-
ties of the metal.

Preliminary absorption isotherms for oxygen in
columbium at 950 and 1050°C show that columbium
rapidly absorbs oxygen in the pressure range from
10 to 1000 i Hg. The absorption rate decreases
as the pressure is decreased. In the pressure
range between 40 and 800 ;. Hg, the formation of
an oxide film competes with the absorption of
oxygen by the metal.

YTTRIUM METAL PRODUCTION
T. Hikido W. J. Werner
Alloy Production

Six additional experiments were completed in the
study of the lithium reduction of a mixture of the
fluorides of yttrium, magnesium, and lithium to
produce an yttrium-magnesium alloy. In run L-8
and subsequent runs both the lithium and the
fluoride mixture were transferred in the molten
state into the reduction retort through resistance-
heated lines from their respective storage con-
tainers by means of differential gas pressures.
At the present time the transfer lines are con-
structed of Z-in.-OD, 25-mil-wall Inconel tubing.
The direct transfer of the molten fluoride mixture
from the fluorinating apparatus to the reduction
retort is now being considered. When this change
is made a 3/8-in. transfer line will be used. This
will facilitate the handling of larger quantities of
material, and larger quantities of alloy will be
produced in each run. A schematic diagram of
the apparatus now being used is presented in
Fig. 1.1.1. A distillation apparatus capable of

handling 4 to 5 |Ib of alloy is used in conjunction
with this apparatus. The ytrium sponge produced
by distillation is arc-melted under an argon at-
mosphere with a tungsten electrode to form finger

 
PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
ORNL —LR-DWG 30443

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

——
-
| - N
. — TRANSFER TUBE
ARGON OUT ~a—
(z--wacuum AND ARGON IN
THERMOCOUPLE TUBE — ] PRESSURE PRESSURE
dll bPAE 4 : 7% TRANSFER TUBE —m
\/ N \\ f F—
NN
/) Z
% 7
% WATER JACKET
—
UL L Z
z .
LITHIUM POT SALT POT
£
)
g ]
%
R THIN - WALLED TANTALUM CRUCIBLE
% ST HEAVY - WALLED TANTALUM LINER
GRAPHITE LINER
Zz
v 7 %

 

 

 

 

Fig. 1.1.1. Diagram of Reduction Retort for the Preparation of Yttrium-Magnesium Alloy, (Secret with caption)

castings of the type shown in Fig. 1.1.2. The
results of chemical analyses of several of these
castings are presented in Table 1.1.2.

A new method for consolidating the alloy
directly into a finished ingot is being investigated.
The electron-beam melting method developed at
Temescal Metallurgical Company is being con-
sidered for this application. In the near future
several ingots of yttrium-magnesium alloy will
be sent to Temescal for melting by this process
and will be returned for evaluation. If the process
proves to be satisfactory for melting this alloy,
the wvacuum distillation and arc-melting steps
can be eliminated.

Evaluation of Material

Tensile Tests. — A series of specimens were
machined from finger castings and pulled at room
temperature. The purpose of these tests was to
see whether there was any direct correlation
between tensile properties and impurity content.
However, as can be seen from Fig. 1.1.2, the
grains in the as-cast specimens are quite large
and nonuniform. A program for determining the
recrystallization properties is being initiated in
order to develop means for controlling the grain
size. It is believed that, when the size of the
grains of the tensile specimens can be con-
trolled, a correlation between impurity content
and tensile properties will be found.

 
ANP PROJECT PROGRESS REPORT

UNCL ASSIFIED
Y-24851

 

040 in./ DIV
Y 5 Y S

Figs 1.1.2. Yttrium Finger Castings L-8A-1, L-8A-2,
L-8A-3. (Secret with caption)

Metallographic Techniques. — In the past the
metallography of yttrium has been hampered by
the lack of adequate polishing and etching
technigues, but a suitable polishing and etching
technique for yttrium has now been developed.®
The specimen is ground dry on 320, 400, and

 

SDevalopud by M. D. Allen of the Metallography Sec-
fiflfll

Table 1.1.2. Resulis of Chemical Analyses
of Castings of Yttrium Sponge

 

 

Run Oxygen Content Nitrogen Content

Number (ppm) (ppm)

L-8A-1 640 710
L-BA-2 1700 1,300
L-8A-3 1900 8,900
L-BA-4 3000 17,000
L-9A 1600 320
L-9B-1 1300 1,100
L-9B-2 1400 570
L-9B-3 1600 1,700
L-9B-4 1800 2,000
L-9B-5 1200 1,400
L-10A-1 610 140
L-10A-2 940 110
L-10A-4 1100 120
L-10A-6 1100 140
L-10A-7 980 100
L-10A-8 850 100
L-11A-1 1200 58

 

600 papers and then ground on 2.0 paper with
a petroleum solvent and paraffin as a lubricant
in the rough polishing step. It is then polished
on a diomond wheel with 20- to 40-u diamond
paste and canvas cloth, 6- to 8- diamond paste
and canvas cloth, and 0 to 1-p¢ diamond paste
and “‘airplane’” cloth. For the final polishing
step, a Syntron vibratory polisher is used, with
(1) Linde A on micro cloth and alcohol plus 10%
oxalic acid as the lubricant, (2) Linde B on micro
cloth and vacuum-diffusion-pump oil as the
lubricant, and finally (3) magnesium oxide on
micro cloth ond vacuum-diffusion-pump oil as the
lubricant. The etchant used is an equal-volume
mixture of HNOa, H,PO,, and CH,COOH. The
etchant is cooled with ice in order to assure a
slow etching reaction, and the sample is sub-
merged in the cold etchant for 3 to 5 sec or until
a thick film is formed on the surface of the
specimen. The sample is then immediately washed
with hot water. The reproducibility of this
procedure is illustrated in Figs. 1.1.3, 1.1.4,

 

 
 

PERIOD ENDING MARCH 31, 1958

 

 

 

Y-25111
w
- L
X
o
z
.
|o.03
Ee3
s O
o
Figs 1.1.3. Sample L-10A-2 Under Bright Field lllumination. 100X. Etchant: HNOB-H3F04-CH3CD'DH-
| UNCLASSIFIED
Y-25112
» Fy 0
- 1j—
x
Q
=
008
I IE.'!
>
-o—
e

 

 

 

Figs. 1.1.4. Sample L-10A-2 Under Polarized Light lllumination. 100X. Etchant: HN03-H3PD4-CH3CUOH-

 
ANP PROJECT PROGRESS REPORT

and 1.1.5. All three figures are photomicrographs
of the same area of sample No. L-10A-2 after
30% reduction by swaging and annealing at 800°C
for 30 min. Figure 1.1.3 is a bright-field photo-
micrograph; Fig. 1.1.4 is a polarized light photo-
micrograph; and Fig. 1.1.5 is a sensitive tint
photomicrograph.

 

Yttrium Hydriding

Several pieces of hydriding equipment were
received from Wright Air Development Center. The
equipment consists of two hydriding furnaces with
associated gas metering tanks, a dissociation-
pressure-measuring furnace, a modulus of elas-
ticity rig, and a thermal stress rig. A program
of study of hydrided yttrium is to be initiated.

 

 

UNCLASSIFIED |
Y-25113 |
; 1 -
~ ]
-li.l""
=z
Qo
=
0.02
0.03
»
- O
o

 

 

Fig. 1.1.5. Sample L-10A-2 Under Sensitive Tint lllumination. 100X. Etchant: HNOSHSPOA'CHECDDH‘

 
PERIOD ENDING MARCH 31, 1958

1.2. CORROSION
E. E. Hoffman

W. H. Cook

D. H, Jansen

Metallurgy Division

LITHIUM PURIFICATION: METHODS OF
REDUCING THE NITROGEN AND
OXYGEN CONTENT

Various means for purifying lithium, including
low-temperature filtration, vacuum distillation, and
gettering with active refractory metals, are being
studied. Commercially produced lithium vsually
contains oxygen and nitrogen in quantities of
several hundred to several thousand parts per
and techniques for reducing the total
oxygen and nitrogen impurity content to less
than 50 ppm are being sought. Lithium as pure
as it is practical to produce is needed (1) for
corrosion tests of container metals in which the
effects of minor amounts of oxide and nitride
can be determined and (2) for use as feed material
in the production of low-oxygen-content yttrium
metal,

million,

Solubility of Nitrogen and Oxygen in Lithium

In order to determine whether cold trapping and
low-temperature  filtration would be practical
methods for reducing the nitrogen and oxygen con-
centration of lithium, several experiments were
conducted to obtain solubility data. Preliminary
data were obtained first on the rate of solution
and the equilibrium solubility of lithium nitride

UNCLASSIFIED
ORNL—LR— DWG 30414

THERMQCOUPLE PROBE
(IRON)

  
  

= F ACE MUFFLE
Mo SAMPLING URNAC

BUCKET
AND BAIL ——

 

 

 

Mo CRUCIBLE

Fig. 1.2.1. Apparatus for Measuring the Solubility of
LigN in Lithium in a Purified Argon Atmosphere.

in molten lithium at 752°F., The test was con-
ducted, in a purified argon atmosphere, in the
apparatus shown in Fig. 1.2,1. Specially prepared
lithium nitride was fused in the bottom of the
molybdenum crucible and allowed to solidify
prior to the introduction of the lithium metal.
The lithium nitride was fused prior to the test
to reduce the pickup of particulate lithium nitride
during sampling. The results of the test, as
tabulated below, indicate that the solubility of
nitrogen at 752°F (400°F above the melting point
of lithium) is quite high:

Time Nitrogen
(min) (wt %)

0 0.11
30 0.99
60 1.50
180 2.10
240 2,86
300 3.14

In order to reduce the possibility of trapping
impurity particles in the samples and yet obtain
a large amount of sample for more complete
analyses, a more refined solubility test rig was
designed and built. This stainless steel apparatus
is shown in Fig. 1.2.2. The test unit was loaded
in a purified inert atmosphere with 60 g of lithium
metal and sufficient lithium nitride or oxide to
give a total oxygen or nitrogen content of 5 wt %.
The system was sealed under 1 atm of pressure
and placed in a furnace. At the end of the test
run, the system was evacuated to a pressure of
less than 5 p. The test unit was then tilted,
as shown in Fig. 1.2.2, to submerge the filter
head. Argon gas pressure was then applied to
the system to force lithium into the evacuated
tube, which was also at the test
temperature. In order to avoid segregation of
the impurities during solidification, the sampling
tube was plunged into an oil—dry-ice bath. The
stainless steel sampling tube was replaced with

sampling
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 24830

VACUUM

   
  
 
 
   

SYSTEM SHOWN TILTED APPROX.
30 deg TO HORIZONTAL BEFORE
PRESSURIZING LITHIUM INTO
SAMPLING TUBE

   

AN
»

 

 

 

 

 

 

   
 

 

 

Iy
\
E
| >
| |
i 5 20- 4 STAINLESS STEEL FILTER
4]
: &
4in 3
| |
‘ N
N
i
i
§ooil
\\‘s
j,// Mi,0 OR LigN [ ~—— SAMPLING TUBE
Fig. 1.2.2. Stainless Steel Test Rig for Determining

the Solubility of Oxides ond Nitrides in Lithium.

one made of titanium for tests of oxygen solubility
to facilitate activation analysis. The results
of these tests are given in Table 1.2.1. These
preliminary data indicate that low-temperature
filtration and cold trapping are not promising
methods of lithium purification if the nitride or
oxide impurity content is to be reduced to below
several hundred parts per million.

Fiftration

Early filtering experiments conducted by GE-
ANPD indicated, in agreement with the conclusion
above, that low-temperature filtration would not
reduce the nitrogen and oxygen content.?2 Also,

 

]G.Leddicotte, Anal. Chem. Ann. Prog. Rep. Dec. 31,
1957, ORNL-2453, p 30.

2A. R. Crocker et al., Design, Operation, and Test of
A Sodium-to-Lithium-to-Air Heat Transfer System, APEX-
178, p 112; APEX-327, p 117 (Dec. 1954).

10

the Nuclear Development Corporation of America
(NDA) has reported unsuccessful attempts to
purify lithium by filtration at temperatures only
slightly above the melting point.®  Filtration
experiments conducted at ORNL have further
substantiated these findings, as shown in Table
1.2.2. The slight increases in nitrogen and
oxygen content upon filtering may be attributed
either to inhomogeneities in the lithium or to
contamination during or following sampling. It
appears therefore that low-temperature filtration
and cold trapping may be beneficial only in
reducing the nitrogen and oxygen content of
lithium that is very highly contaminated with
nitrogen or oxygen.

Yacuum Distillation

Vacvuum distillation of lithium as a means of
reducing the nitrogen and oxygen content has
been studied at ORNL and at NDA. The NDA
system has been used successfully to reduce
the nitrogen content to below 10 ppm. In the
NDA system, titanium sponge is added to the
stifl prior to distillation to getter the nitrogen
and to assist in the purification. No titanium
was added to the still in the distillation performed
at ORNL. The ORNL distillation system and
allied equipment are shown in Figs. 1.2.3 and
1.2.4. Approximately 1100 g of lithium is charged
to the still and 80% of each charge is distilled
into the receiver. The remaining 20% is dis-
charged through the drain line at the completion
of each distillation. The distillation rates average
30 g/hr or 2.5 g/hr per square inch of liquid-vapor
interface area.

Six distillations have been performed to date,
and the nitrogen and oxygen content of the
distillate has been of the same magnitude as or
higher than that of the lithium charged into the
still in all distillations. The results of the six
distillations are given in Table 1.2.3. The
system operated in a satisfactory manner except
as noted in the table. The operating conditions
were identical with those used at NDA, insofar
as possible, in order to ascertain the effectiveness
of distillation when no titanium getter was added
to the still. The results indicate that distillation
without gettering is not a practical means of

 

3W. Arbiter and S. Lazerus, Purification of Lithium by
Vacuum Distillation, NDA-39, p 7 (June 14, 1957).

 
PERIOD ENDING MARCH 31, 1958

Table 1,2,1, Solubility of Nitrogen and Oxygen in Molten Lithium

 

 

Filtering Aging Time Prior
Test Solute* Temperature to Sampling Sample  Solubility Remarks
NO. (OC) (hr) No- (ppm)
1 Li3N 300 64 1 4,060 Samples slowly cooled; considerable
2 7,560 segregation occurred
3 19,600
2 LizN 250 24 1 1,000 Samples quenched to prevent
2 920 segregation
3 Li3N 250 72 1 1,080 Samples quenched
2 1,380
3 1,400
4 Li3N 300 24 1 1,760 Samples quenched
2 1,900
5 Li20 250 24 1 19,700 Samples quenched
2 22,700

 

*The test chamber was loaded with 60 g of |ithium metal

lithium used contained 1076 ppm N, and 550 ppm 02.

Table 1,2,2, Results of Tests of Low-Temperature

Filtration as a Means of Purifying Lithium

 

Impurities? (ppm)

 

 

02 N2 C
As-received |ifhiumb
Sample 1 334 560 15
2 446 645
Filtered tithium®
Sample 1 595 680 11
2 518 760 18

 

2Analytical methods: (1) activation analysis for oxe
ygen, (2) steam distillation followed by colorimetric
analysis for nitrogen, and (3) spectrophotometric deter«
mination of lithium carbide in metallic lithium as the
acetylene—~silver perchlorate complex.

Packed by vendor in stainless steel gastight cone
tainers under helium.

“Transferred through S-u stainless steel filter at 300°C
following aging at 300°C.

and 5 wt % nitrogen (as Li3N) or oxygen (as Li20). The

purification. Lithium samples from all the purifi-
cation tests are being exchanged between NDA
and ORNL in order to check the reproducibility of
the analytical results at the two installations.

Gettering with Zirconium, Titanium, and Calcium

The results of the filtration, distillation, and
solubility tests described above indicated that
gettering of the nitrogen and oxygen by active
metals might be the best method of purification.
The free energies of formation of various oxide
and nitride compounds of interest were therefore
examined (Table 1.2.4) in order to select the
gettering metals to be tested. As may be seen,
a considerable number of metals would probably
getter nitrogen from lithium metal, but the oxide
of lithium is extremely stable and only a few
metals, among them calcium, yttrium, thorium,
and beryllium, have oxides whose free-energy
values indicate that they might be suitable as
getters for reducing the oxygen content of lithium.
Studies of the purification of lithium by gettering
have thus far been limited to tests with zirconium,
titanium, and calcium,

11
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 22264 R
FULTON

BELLOWS VAL\!"E] PRESSURE GAGE

   

_AVACUUM LINE { Y4-in. TUBE)

 

 

  
   
  
   
 
 
 
    
  
    
 

PRESSURE LINE

__—6-in. FLANGE AND O-RING
—g-in. TUBE
-MICRO METALLIC FILTER 104
£, _—Yo-in. SCH 40 PIPE

i_, 2

—THERMOCOUPLE GLAND

DISCHARGE LINE
TO STILL

  

of :

GAS FILL LINE

———— ADJUSTABLE LIQUID LEVEL PROBES

     
  
   

L

GAS
RESERVOIR

    
  
 
 
   
   
  
  
   

110w TO AUTOMATIC SAFETY

___-SOLENOID VALVE (NORMALLY CLOSED)

NOTE!
VACUUM LINES ARE 1¥in. COPPER TUBING ;
ALL OTHER MATERIAL TYPE 347 STAINLESS
STEEL. VACUUM SYSTEM VALVES ARE 13g-in.
BELLOWS-TYPE VEECO.

 

FILTER Tt i ALPHATRON
FILTER TANK TANK FiLL 19g-in. VEECO o J
VALVE IBRASSy VACUUM GAGE
SQ0-WATT i HASTINGS
ELEMENT THERMOCOUPLE
GAS- .
. I"‘\ n.‘l
o '
570 |
FURNACE — lasoec) ~ADJUSTABLE LIQUID

   
   
 

— i
£
4-in, ISOTHERMAL |-0] |
SODIUM JACKET — 7|8

3-in. BOILER POT—+H+

CLAM SHELL
HEATER— |

t-in. CONDENSER ,;"

 

THERMOCOUPLE (675°C)- NE 4 (250 WATTS EACH) coLD TRAP =
. LI i n
DRAIN LINE- 3-in. RECEIVER P't:lT-’"gfi

     
    
     

i
i

o

A

 

LEVEL PROBES (2)
THERMOCOUPLE GLAND
T GAS

VAPOR BAFFLE

-4 STRIP HEATERS

~THERMOCOUPLE (190°C)

    

   

 

[}

M.C.F.

CENCO-HYPERVAC—25

- T AGE TANK
DRAIN LINE TO STORAGE S ROUGHING PUMP

DIFFUSION PUMP

Fig. 1.2.3. Diagram of Lithium Distillation System.

 

STILL
— FURNACE

UNCLASSIFIED
0

      
 
   
    

UNCLASSIFIED
¥Y=-23T78

CONDENSER

RECEIVER

Fig. 1.2.4. Lithium Distillation System With and Without Furnaces and Insulation.

12

 
PERIOD ENDING MARCH 31, 1958

Table 1,2,3, Effect of Vacuum Distillation at 650°C on the Nitrogen and Oxygen Content of Lithium

Yacuum at cold trap with system cold: 2 x 10~ mm Hg
Vacuum at cold trap with system ot temperature: 1 X 104 mm Hg

 

Distillation No. Nitrogen* (ppm) Oxygen* {ppm) Remorks

 

Charge to still (1100 g) 1170, 1250 226 As received and filtered ot
300°C through a 5-p filter
1 1400, 1300 370, 285
Av 1350 Av 327
2 1260, 1150, 885, 1010 700, 389, 376, 533
Av 1076 Av 500
3 3000, 1900 375 New still (4 in, pipe); condenser
Av 2450 leaked near end of run
4 4200, 5300 1120, 1210 Some contamination of system occurred
Av 4750 Av 1165 during repair of leak following dis-
titlation No. 4
5 1600, 1600, 2000 203, 198
Av 1733 Av 200
6 1800, 2100, 2700, 3100 701, 758, 688, 630,
Av 2425 Av 694

 

*Each value is for a separate analysis or is an average, as indicated, of the values given,

Zirconlum Gettering Test. — Only one gettering
experiment with zirconium has been conducted
thus far. The appearance of the 0.022-in.-thick
zirconium getter sheet before and after the test
may be seen in Fig. 1.2.5, and the test results
are summarized in Table 1,2.5. Metallographic
sections of the zirconium specimen are shown
in Fig. 1.2.6. The results of chemical analyses
of the lithium and the zirconium indicate that
zirconium is an efficient nitrogen getter under
the conditions of this test. The absence of an
increase in the oxygen content of the zirconium
is in agreement with the free-energy data given
in Table 1.2.4. The zirconium sheet weighed
33.913 g before the test and 33.933 g after the
test; that is, there was a net weight gain of
0.65 mg/in.2.

Titanium Gettering Tests. — Three lithium purifi-
cation tests were run with titanium foil as the
gettering and two tests were run with
titanium sponge. It was realized that much more
efficient gettering would be obtained with sponge,
because of the larger surface area per unit of
but foil was used in the preliminary

agent,

weight,

experiments so that precise data on the gettering
rates and the formation of oxide or nitride layers
could be obtained by x-ray and metallographic
examinations.

The conditions of the first two tests with
titanium foil were similar except for the test
which was 1400°F in one test and
1500°F in the other. The test conditions and
results are given in Tables 1.2.6 and 1.2.7.
The results of both experiments show that titanium
will getter nitrogen from lithium quite rapidly at
both 1400 and 1500°F. The rate of nitrogen
removal from the lithium was highest in the first
24-hr test period at 1500°F, as would be expected,
since the diffusion rate of nitrogen into the
titanium is greater at the higher temperature.
The considerable decrease in the rate of removal
during the second 24-hr test period emphasizes
the need for a high ratio of titanium surface to
lithium volume. As would be expected from the
free-energy data, titanium is not effective in
gettering oxygen from lithium. The considerable
in the oxygen content of the [ithium

temperature,

increase

is not understood; but, since the analytical

13

 

 
ANP PROJECT PROGRESS REPORT

Table 1.2.4. Free Energies of Formation of Various Oxide and Nitride Compounds of Metals

of Interest as Getters in the Purification of Lithium

 

—AF® (kcal per gram atom of nonmetal)

 

 

Metal Compound -
At 25°C At 223°C At 727°C
Lithiom Lizo(“) 138 130 123 -
LioN(®) 37
Calcium ca0fc) 144 139 128
1 (5)
Yttriom % ¥,0,(4) 143
yN(®) 64
Thorium % Tho, () 140 136 125
] (b)
Beryllium Be0(©) 139 134 123
] (b)
BeaN, 61
Uranium %u0,() 128 125 117
un(®) 75 .
Barium Ba0(c) 126 120 107
i b)
Y BagN, 37
Zircon ] 1
irconium /22r02 124 119 108
zeN(D 75 71 60
Titanium %y Ti0, 8 101 98 88
TiN(®) 73
Columbium I/Cb 0 (c) 85
5 275
CbN(®) 53
1 (c)
Molybdenum 5Mo0, 60
Mo,N(6) 10

 

(@) E. Weber and L. F. Epstein, J. Met. and Ceramics, Issve No. 3, TID-67, p 85-110 (May 1949).

(b)L. Brewer et al., Natl. Nuclear Energy Ser. Div IV 19B, 44 ff. (1950). )
(C)F. D. Rossini et al., Natl. Bur. Standards(U.S.) Circ. 500 (Feb. 1, 1952).

(d)i. E. Campbell (ed.), High-Temperature Technology, Wiley, New York, 1956.

]95:6)0. Kubaschewski and E. L. Evans (eds.), Metallurgical Thermo Chemisiry, vol |, Academic Press, New York,

(f/)B. Lustman and F. Kerze, Jr. (eds.), The Metallurgy of Zirconium, McGraw-Hill, New York, 1955.
(S)F. D. Richardson and J. H. E. leffes, |. Iron Steel Inst. (London) 160, 261 (1948).

14

 
PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
Y-24728

 

0L\ 1] (o]0

  

Fig. 1.2.5. Zirconium Sheet (a) Before and (b) After Exposure to Lithium at 1400°F for 24 hr. Surface colors
on the zirconium specimen after the test were bronze and purple.

UNCLASSIFIED

 

 

Y-25514
'rs:nx E ]E E ml:HE:s Table 1.2.5. Results of Lithium-Zirconium Gettering
= T T : Test Ne. 1
& ) 2 Volume of lithium: 23.4 in.>
B

Weight of lithium: 180 g

Surface area of zirconium sheet: 30.6 in.?

Ratio of zirconium surface arec to lithium
volume: 1.31in.%/in.3

Container: Armco iron

 

™
!
N
i o
i T

 

-~ ..'I' ; . Test temperature: 1400°F
. fl{-' .
i .é.' = ' PR, (Hp e . . Impurity Content
: E ZrN D : X
Yo G.z'__ b LG 2 it (ppm)
5P aterial Analyze
= ok # t\“‘ 2 02
BN OF N (
L Lithium
Before test 1100
SURFACE BEFORE "|"| SURFACE AFTER
EXPOSURE EXPOSURE After 24 hr 630
Zirconium
Fig. 1.2.6. Metallographic Sections of Zirconium Spec-
imens (@) Before ond (k) After Exposure to Lithium for Before test 70 1000
24 hr ot 1400°F. Note 0.0002-in. layer of Z¢rN on surface After 24 hr 510 1000

of exposed specimen. As polished. 750X, Reduced 28%.

 

15

 

 
ANP PROJECT PROGRESS REPORT

Table 1.2.6, Results of Lithium=-Titanium Gettering Test No. 1

Test temperature: 1400°F
Volume of lithium: 103.9 in.3
Weight of lithium: 800 g

Surface area of titanium foil; 280 in.2

Ratio of titanium surface area to lithium volume: 2.69 in.z/in.a

Container: type 347 stainless steel

Weight of titanium foil:
Before test: 128.910 g
After test: 129.578 g

Net gain: 0.668 g or 2,38 mg/in.2

 

Impurity Content

 

Material Analyzed (ppm) Remarks
N, O,
Lithium
Before test 3100 200
After 24 hr 1600 1230 Rate of N2 loss (0—24 hr), 62 ppm/hr
After 48 hr 900 2280 Rate of N, loss (24—-48 hr), 28 ppm/hr
Titanium foil
(0.012 in. thick)
Before test 73 1500
After 24 hr 1800 1400 X-ray examination of surface showed Ti, Ti{N),
and TiN; metallographic examination showed
0.0002-in. layer of TiN
After 48 hr 2000 1500 Xeray examination of surface showed Ti, Ti(N),

and TiN; metal lographic examination showed

0.0004+in, layer of TiN

 

procedure {activation analysis}) employed for
determining the oxygen content is difficult and
segregation of impurities in the test samples is
a problem, it is possible that the increase in
the oxygen content is an analysis error. Material
balances indicate that the increased nitrogen
content of the titanium accounts for only approxi-
mately 40% of the observed weight increase of
the foil. The surface color of the titanium was
bronze following 48 hr of exposure to lithium
at 1400°F and indigo following 48 hr of exposure
at 1500°F. Polished and etched cross sections
of the titanium specimens following the various
tests are shown in Fig. 1.2.7.

In test No. 3, which was run at 1500°F for

a total of 192 hr, the stainless gettering container

16

was loaded with 280 in.Z of titanium foil that
remained in the lithium for the entire test. This
foil was wrapped on a stainless steel holder as
in Fig. 1.2.8. Small titanium tabs with
6-in.?2 surface areas were inserted into the molten

shown

lithium during certain time periods of the test
run as given in Table 1.2.8. The purpose of
this was to determine the effectiveness of a
fresh getter surface as the nitrogen content of
the lithium decreased. The analytical data
obtained from this test are presented in Tables
1.2.8 and 1.2.9. As may be seen, the nitrogen
content of the lithium was lowered significantly
during the experiment, and the rate of nitrogen
pickup by fresh titanium did not
appreciably as the nitrogen content of the lithium

decrease

 

 
PERIOD ENDING MARCH 31, 1958

Table 1.2.7. Results of Lithium=Titanium Gettering Test No. 2

Test temperature: 1500°F
Volume of lithium: 94.8 in.®
Weight of lithium: 730 g
Surface area of titanium foil: 283 in.2
Ratio of titanium surface area to lithium volume: 2.98 in.zfin.a
Container: type 347 stainless steel
Weight of titanium foil:
Before test: 129,287 g
After test: 130.292 g

Net gain: 1.005 g or 3.55 l'rlg,,i"in..2

 

Impurity Content

 

Material Analyzed (ppm) Remarks
Ny 0y
Lithium
Before test 2450 370
After 24 hr 740 910 Rate of N2 loss (0=24 hr), 71 ppm/hr
After 48 hr 380 730 Rate of N, loss (24-48 hr), 15 ppm/hr

Titanium foil
(0.012 in. thick)

Before test 73 1500

After 24 hr 1600 1000 Xeray examination of surface showed Ti(N) and TiN;
metallographic examination showed 0,0004-in.
layer of TiN

After 48 hr 3200 1400 X-ray examination of surface showed Ti(N) and TiN;
metallographic examination showed 0.0006~in,
layer of TiN

 

UNCLASSIFIED
Y-25M3

 

BEFORE 24 hr AT 1400°F 48hr AT 1400°F 24 hr AT 1500°F 48 hr AT 1500°F

Fig. 1.2.7. Surfaces of Titanium Gettering Specimens Before and After Exposure to Molten Lithium. Exposed
surface at bottom. Note titanium nitride layer. Etchant: HF-HNO,-glyceria. Reduced 28%.

17

 
ANP PROJECT PROGRESS REPORT

dropped from 550 to 100 ppm. At some nitrogen
content in the lithium of less than 100 ppm very
little titanium nitride formed on the fresh titanium
surface (see 120 to 192 hr specimen), and, instead,
massive quantities of nickel were picked up by
the titanium from the lithium. Titanium tabs
inserted at various periods are shown in Fig.

UNCLASSIFIED
Y-25444

.

 

Fig. 1.2.8. Titanium Gettering Foil ond Stainless
Steel Support Assembly for Foil.

 

1.2.9; the structure change resulting from nickel
alloying may be seen. |t appears therefore that
stainless steel is probably not a satisfactory
container material for purification tests. The
nickel, iron, and chromium content of the lithium
samples was lower than expected at the 1500°F
sampling temperature. |t has been reported that

Table 1.2.8. Analyses of Lithium Samples from
Lithium-Titanium Gettering Test No. 3

Test temperature: 1500°F

Volume of lithium: 102.3 in.3

Weight of lithium: 788 g

Surface area of titanium foil: 280 in.2

Ratio of titanium surface area to lithium
volume: 2.73 in.z.a*’in.3

Container: type 347 stainless steel

 

Impurity Content (ppm)

 

 

Lithium

Samples N, 0, Fe Ni Cr
Before test 1400
After 24 hr 550
After 48 hr 350 1080 40 90 49
After 72 hr 200 2790 32 70 23
After 96 hr 100 1750 61 58 47
After 120 hr 125 694 29 65 13
After 192 hr 53 675

 

 

 

Table 1.2.9. Analyses of Titanium Samples from Lithium-Titanium Gettering Test No. 3

 

Impurity Content (wt %)

 

Weight Gain of

 

Titonium Samples Foil During
N, 0, Fe Ni Cr Exposure (mg/in.

Before test 0.044 0.14 0.17 0.017 0.010

0-24 hr 0.250 0.16 0.19 0.025 0.014 1.8
24-48 hr 0.140 0.17 0.18 0.027 0.012 0.9
48-72 hr 0.180 0.16 0.35 0.039 0.019 1.2
76-92 hr 0.140 0.14 0.16 0.033 0.014 1.3
96-192 hr* 0.200 0.17 0.21] 0.90 0.022 12.4
120-192 hr** 0.061 0.16 0.25 10.10 0.117 57.9
0-192 hr 0.3%90 0.15 0.26 0.23 0.0M 4.9

 

*Note high nickel content and large weight gain during exposure period.

**Note low nitrogen content, high nickel content, high chromium content, and large weight gain.

18
PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
¥- 25451

 

ANALYSIS OF SPECIMEN
CROSS SECTION (wt %)

N-0.25
0—-0.16
Ni—=0.02
Cr—0.01
Ti— BAL

 

‘

 

 

N—0.06
0- 0.16
Ni—40.10
Cr—0.42
Ti —BAL

 

4

b
B

o
- e ,I:l’ ??;'—l*;\
2 Woh sl
0.012 in.

 

 

N-0.39
0—-0.15
Ni—-0.23
Cr—0.01
Ti—BAL

 

Fig. 1.2.9. Titonium Specimens from Lithium-Titanium Gettering Test No. 3. (2) Tab exposed during 0- to
24-hr test peried. (b) Tab exposed during 120- to 192-hr test period. Note extensive phase changes as a result of

high nickel pickup. (c) Specimen exposed during 0- to 192-hr test perioed. Etchant: 25% HF -25% HN03—5D%
glyceria. Reduced 20%.

19

 
ANP PROJECT PROGRESS REPORT

the solubility of nickel in lithium is 1580 ppm
at 1200°F.*

Analyses of the stainless steel container showed

(Table 1.2.10) that stainless steel does not
contribute  significant amounts of nitrogen to
the fithium. There was some mass transfer of

the titanium to the stainless steel container

wall.

Table 1.2.10. Results of Analyses of Type 347
Stainiess Steel Container Used in Lithium-T itanium

Gettering Test No. 3

 

Impurity Content {(wt %)

 

Description of Sample

 

02 N2 Ti
Before test 0.048 0.086 0.032
After test
0.050-in. cut from 0.045 0.074* 0.055
inside wall
0.050-in. cut from 0.029 0.038*

ocutside wall

 

*Oxide scale on outside removed before sample taken
for analysis.

Approximately 20 kg of badly contaminated

lithium was recently received from Nuclear
Development Corporation of America. This lithium
had been exposed to air in 2-Ib lots for a consid-
erable period of time. Of this material, 10.4 kg
was pressurized through a 65-u stainless steel
filter at 250°F into a stainless steel container.
A 4-in.-dia high-temperature finger containing
1.7 kg of titanium sponge extended from the
bottom to the top of the 12-in.-dia stainless steel
container,

The purification tank, the lithium, and the
hot finger were maintained for 100 hr at T000°F
and then for another 100 hr at 1400°F. The
nitrogen content of the lithium after filtering
and before gettering was 2250 ppm and after
the gettering was 33 ppm. The oxygen content,
which was 950 ppm after gettering, was not
substantially altered by the purification treatment.

Lithium-Titanium-Calcium Gettering Test. — As

indicated above, the purification of lithium by

 

4. Q. Bogley and K. R. Montgomery, The Solubility
of Nickel in Lithium, IGR-TN/C-250 (Sept. 30, 1955).

20

gettering the nitrogen and oxygen with titanium
has only been partially successful. Titanium
has been an excellent getter for nitrogen in all
tests thus far, but it seems doubtful that it will
be wuseful in reducing the oxygen content of
lithium. In an effort to prepare low-oxygen-content
lithium for use in yttrium metal production, gettering
with both titanium metal sponge and triple-distilled
calcium was tried. Calcium is one of the few
metals whose oxide has a lower free energy than
that of lithium oxide (see Table 1.2.4). There
is considerable disagreement between the two
available calcium-lithium phase diagrams, par-
ticularly with regard to the solubility of calcium
in lithium at temperature slightly above the
melting point of lithium, At 400°F for instance,
Zamotorin® gives the solubility of calcium in
liquid lithium as approximately 3%, while Wolfson®
gives a value of approximately 60%.

The conditions and results of the first test
with titanium ond calcium are given in Fig.
1.2.10. As may be seen, the purification procedure
reduced the nitrogen content to a very low value
but the oxygen content actually increased. The
fairly low calcium content (1.95%) of the lithium
product as compared with the amount (10%) placed
in tank No. 1 indicates either that the solubility
of calcium in liquid lithium is lower than the
valué given by Wolfson or that the calcium alloyed
with the titanium in tank No. 2 during the high-
temperature gettering treatment, The latter pos-
sibility is thought to be the more probable.

STATIC LITHIUM CORROSION OF YTTRIUM
E. E. Hoffman

An yttrium metal specimen was exposed to
static lithium for 100 hr at 1500°F in a columbium
container tube. The test assembly and the con-
siderable amount of yttrium metal which mass-
transferred from the test specimen to the inner
wall of the columbium tube are shown in Fig.
1.2.11. The yttrium cylinder decreased in diameter
from 0.210 in. to as low as 0.207 in. in some
areas, and it lost 0.10 g/in.2, or 2.4%, in weight.
The extensive attack of the yttrium is illustrated

 

M. I. Zamotorin, Metallurg 13(1), 96—99 (1938).
6M. R. Wolfson, Trans. Am. Soc. Metals 49, 794 (1957).

 
LITHIUM CHARGE
840 g

Oo,—= 730 ppm
No — 2900 ppm

 

 

 

PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
ORNL—LR—-DWG 31097

GETTERED LITHIUM
0, —1040 ppm

N, —102 ppm

—1.95%
LITHIUM SAMPLING TUBE ——"fl-——|_—C° 95%

 

 

 

 

 

 

 

 

 

TANK NO.

 

TANK NO. 2

TANK NO. 3

 

 

 

 

 

 

Z-—:—— 84 g Ca \

N, 240 g Ti
S SPONGE

 

 

 

 

 

 

 

 

 

CHARGED WITH LITHIUM

HELD AT 1200°F -1 hr

PRESSURIZED INTO TANK
NO. 2 AT 750°F

CHARGED WiTH CALCIUM—
LITHIUM ALLOY

HELD AT 1400°F ~12 hr

HELD AT 1000°F —80 hr

 

STORAGE TANK

PRESSURIZED INTO TANK
NO. 3 AT 600°F

Fig. 1.2.10. Test Conditions and Results of Lithium-Calcium-Titanium Gettering Test No. 1.

in Fig. 1.2.12, which shows the exposed end
of the yttrium cylinder following the test. The
specimen,
test, was heavily etched by the lithium, par-
ticularly along the grain boundaries. [t appears
therefore that the solubility of yttrium in lithium
at the test temperature of 1500°F is appreciable.

A metallographic section of the yttrium specimen
is shown in Fig. 1.2.13. A phase, which has
tentatively been identified as yttrium oxide,
may be seen along the surface and in the grain
boundaries. The bulk oxygen content of the
yttrium specimen increased from 1900 ppm before
the exposure to lithium to 3000 ppm following
test. The possibility of using yttrium metal as
a gettering agent to reduce the nitrogen and
particularly the oxygen concentration of the
lithium to be used in the subsequent production
of low-oxygen-content yttrium metal is being
considered.

which was machine-ground prior to

STATIC LITHIUM CORROSION OF
ZIRCONIUM-BASE ALLOYS

E. E. Hoffman

Difficulty has been encountered
ductile saddle

the use of columbium weld-filler rod.

in making
columbium welds with

Therefore
two zirconium-base alloys were tested for use
as welding-rod alloys in the fabrication of various
columbium corrosion test components. Both an
85% Zr-15% Cb and an 80% Zr-15% Cb-5% Mo
alloy were used, and the binary alloy was found
to yield the most ductile, crack-resistant weld-
ments. The details of these welding studies
have been reported,” and the results of static
corrosion tests are given below. In addition,

tubing

 

’R. L. Heestand and E. E. Hoffman, Welding of
Columbium. Part 1, ORNL-2423 (to be published).

21

 
 

ANP PROJECT PROGRESS REPORT

‘ COLUMBIUM TUBE

 

LITHIUM LIQUID LEVEL

 

 

 

1 ‘.-—- YTTRIUM SPECIMEN

 

 

Fig. 1.2.11.
1500°F. Reduced 18%.

 

Fig. 1.2.12. Surface of the End of an Yttrium Cylinder
Following Exposure to Lithium for 100 hr ot 1500°F.
Reduced 34%. (Secret with caption)

22

   

 

 

‘,'.s - &
gl b
-:;.0 “‘ -"i

- ‘poe

0.25 in,

el e T % ";. Moo v
B Ne ?i:"‘;rfluu €230y alom,

";'a -,

INNER WALL OF COLUMBIUM TUBE

Yttrium-Lithium Corrosion Test Assembly ond Inner Wall of Columbium Container After 100 hr at

several columbium thermal-convection loops have
been fabricated and tested in which the saddle
welds and butt welds were made with the 85%
Zr-15% Cb alloy. During relatively short-time
exposures to lithium in these loop tests (115 hr
at 1500°F and 36 hr at 1600°F), weld joints made
with the binary alloy were found to have excellent
corrosion resistance to the lithium. Details of
these tests are given in the following section.

In the static tests, specimens of the B85%
Zr—15% Cb alloy and the 80% Zr-15% Cb-5% Mo
alloy were exposed to lithium for 100 hr at 1830°F.
The ternary-alloy specimen showed no weight
change as a result of the exposure, and the binary
alloy showed a very slight weight gain (0.001
g/in.?). As may be seen in Fig. 1.2.14, neither
specimen was ottacked. The 85% Zr-15% Cb
alloy was used to seal the columbium corrosion
test capsule, and the oppearance of the weld
joint following the test is shown in Fig. 1.2.15.
Some intergranular cracking occurred near the
top of the weld joint as a result of the thermal
treatment during the test. However, the test
PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
Y-25489 ||

L 2
o

INGI-llE‘B

w " -4 - . . g “- .-“ . .-'. : F - . "N 3 ol ; ) i ; -

2 - 3 g W T e % " s 0 s .l-' : e J "4 . v T

\.l.\- [_F b E * £ - o L <.\ e W s n A - g g - 3 : .WH
*H‘_‘ ‘-L}“-“:"\ N .I - i ‘ ‘__ .. : : _.- i -'-‘-l ': .':: = wh_ . Las & I ] o o I s -: B f .

4; i “.‘. lfik.‘ -r ‘ '

 

T
5?0!

 

 

 

 

Fig. 1.2.13. Metallographic Section of Yttrium Specimen Following Exposure to Lithium for 100 hr ot 1500°F.
White phase along surface (top) and the grain boundaries has tentatively been identified as yttrium oxide. Etchant:
HNDB-HC2H302-H20. 500X. (Secret with caption)

UNCLASSIFIED
¥Y-25070

 

H&\E} é‘ ;- 003

-

~

w
&

2 |fi "INCHES
~ I 1
1\

2

2
-

 

o
o™

80% Zr—15% Cb — 5% Mo

I
250%

1

85% Zr —15% Cb (

 

 

Fig. 1.2.14. Metallogrophic Sections of Two Zirconium-Base Alloys Following Exposure to Lithium for 100 hr
at 1830°F. Cathodic etch. 250X. Reduced 12%. (Secret with caption)

23

 

 
ANP PROJECT PROGRESS REPORT

Zr-15% Cb ¥
% Ty A

(o) 50 %

 

- R

Fig. 1.2.15. Section Through End of Columbium Tube Where 85% Zr—15% Cb Alloy Was Used os Filler Rod To
Seal the Crimped Edges. Specimen was exposed to lithium for 100 hr at 1830°F. Enlarged view of cracked section

in () may be seen in (k). As polished. Reduced 17%. (Secret with caption)

capsule did not fail. Although most of the cracking
occurred in the brittle grain-boundary phase, some
transgranular cracking was also detected. The
diamond pyramid hardness of the Zr-Cb weld
joint was 236 as brazed, and the weld was ductile.
The hardness was essentially the same following
the test, 238 DPH, but the joint was found to be
extremely brittle, as shown in Fig. 1.2.15.
Although no attack was detected in the base

material of the sintered columbium tube wall,
heavy intergranular penetration by lithium was
detected in the longitudinal tube weld and in

the heat-affected zone of the weld. This attack
was found to extend to a depth of approximately

0.025 in.

THERMAL-CONVECTION LOOP TESTS
OF LITHIUM IN COLUMBIUM
E. E. Hoffman

As was reported previously,® three of the four
columbium thermal-convection loops that have

24

circulated lithium have shown heavy intergranular
attack. These earlier tests and the tests to be
discussed here were conducted in a stainless
steel vacuum chamber. The procedure used, as
illustrated in Fig. 1.2.16, permits the testing
of unclad columbium loops. The attainable
vacwum (less than 5 p with the loop at operating
temperature) has been adequate to prevent oxi-
dation or embrittlement of the columbium tube
wall. The tests conducted to date have been
of limited duration because of the high rate of
penetration of the longitudinal tube wall weld
by the lithium metal. The penetration is illustrated
in Fig. 1.2.17(a), which is a picture of a section
of tubing from the hot leg of loop No. 58, and
in Fig. 1.2.17(b), which is a transverse section
through the tube wall at the location of the leak.
Both the loops (Nos. 58 and 59) to be described

 

BE. E. Hoffman and L. R. Trotter, ANP Quar. Prog.
Rep. Dec. 31, 1957, ORNL-2440, p 173.

 
    
  

\
|
A1

CLAM-SHELL
HEATERS

    
    

|
|
|
|
|
l
|
|
i
LOOSELY WRAPPED l
HEATING ELEMENT |
{TURNED OFF AFTER |
FLOW HAS STARTED)'\f f
!
4

 

CRIMPED AND WELDED

 

fi ’ B P

{ LA A7 e
/‘: a5
\ i

SEE SPECIMEN
DETAIL

L~ INSULATION

PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
ORNL-I_LR-DWG 23499

   
    
   

SPECIMEN SUPPORT
WIRE

0.035-in. THICK
SPECIMEN

DOOR —

STAINLESS STEEL
BOX— -

Location of Loop in Vacuum Box.

Fig. 1.2.16. Thermal-Convection Loop in Yacuum Chamber.

here were constructed of I/2-in.-OD, 0.035-in.-wall
sintered columbium tubing and each contained
an insert of arc-cast columbium tubing in the
hot-leg section, as shown in Fig. 1.2.18. The
tests of both loops were terminated by lithium
penetration of the longitudinal tube wall weld.
The extent of the attack in the hot and cold
of these loops is given in Table 1.2,11.

legs
Very little attack was detected in the base
material of the tube wall away from the weld

and the heat-affected zone of the weld. In general,
the intergranular attack in the heat-affected zones
adjacent to weld areas was as deep as in the
the weld zone itself. The metallographic sections
showing the attack of the arc-cast insert specimen
and of the sintered tube wall are presented in
Figs. 1.2.18 and 1,2.19. A hardness survey
(25-g load) was made on the tube wall, as shown

in Fig. 1.2.18, in order to determine whether

there were any areas of embrittlement in the
tube wall which might indicate pickup of elements
such as nitrogen or oxygen either from the lithium
or the atmosphere (vacuum) during test. No such
embrittlement was detected, and, in fact, the
arc-cast specimen was softer following test than
before. The mechanism of the attack detected
primarily in the weld zones and heat-affected
areas of these loops is not known at present,
but it is thought to be due to precipitation or
segregation of impurities into the grain boundaries
during the welding operation. This does not,
however, explain the heavy attack of the arc-cast
columbium tube inserts in both tests. The sheet
specimens that were exposed in the hot legs in
the position shown in Fig. 1.2.16 were examined
for weight change. The specimen in loop 58
showed a weight loss of 7.5 mg/in.?, and the
specimen in loop 59 gained 0.62 mg/in.2.

25

 

 
ANP PROJECT PROGRESS REPORT

  
  

 

i UNCLASSIFIED
Y-24729

 

'UNCLASSIFIED
Y-24788

VACUUM

 

Fig. 1.2.17. Section of Sintered Columbium Tubing from Hot Leg of Loop No. 58 (See Fig. 1.2.18 for Over-All

View).

(@) Columbium tube showing location of lithium penetration through longitudinal weld.

(&) Metallographic

section of tube wall weld zone showing penetration along grain boundery. Etchant: HF-H2504-HN03-H20. Reduced

6%. (Secret with caption)

Nitrogen and oxygen analyses of various sections
of the columbium from loop No. 58 are given in
Table 1.2.12. For these analyses, weld metal
was cut from the tube wall and analyzed separately
from the base material. The nitrogen and oxygen
content was slightly higher in all the specimens
tested than in the as-received material. Static
capsule tests are presently being conducted on
specimens of various grades of columbium sheet
that were inert-gas welded with varying degrees
of protection during welding to study the effect
of pickup of contaminants such as oxygen and
nitrogen on corrosion resistance.

STATIC LITHIUM CORROSION OF COLUMBIUM
E. E. Hoffman

A series of static tests was conducted in order
to determine whether the type of environment
(vacuum, 1 to 5 ) in which the thermal-convection

26

loop tests were conducted was in any way re-
sponsible for the intergranular attack of the weld
and heat-affected zones of the loop material.
Four sintered columbium tubes of the same
material (N,, 0.016%; O,, 0.027%) as that used
to construct loops 58 and 59 were loaded with
lithium metal. Tweo tubes were then tested in
a stainless steel container as illustrated sche-
matically in Fig. 1.2.20(¢), and the other two
tubes were tested in the vacuum chamber as
shown in Fig. 1.2.20(4). The tests were conducted
at 1500°F for 100 hr. Argon gas (160 liters) con-
taining 5 ppm O, and 58 ppm N, was circulated
through the test assembly shown in Fig. 1.2.20(a).
A dynamic vacuum of less than 5 p was maintained
around the test assembly shown in Fig. 1.2.20(b).
At the completion of the tests, various sections
of the test capsules were examined metallo-
graphically. The bath and vapor zones of each
of the four columbium tubes were found to have

 

 
 

Table 1.2.11.

PERIOD ENDING MARCH 31, 1958

Results of Columbium-Lithium Thermal-Convection Loop Tests Nos. 58 and 59

 

 

Loop No. Location of Specimen Metallographic Data E
) 58 Hot leg
Arc-cast insert Attack to a depth af 7 mils
Sintered tube
Weld zone Attack to a depth of 7 mils in most areas,
complete penetration in several areas
Base material No attack
Cold-leg sintered tube
Weld zone Attack to o depth of 5 mils
Base material No attack
59 Hot leg

Arc-cast insert
Sintered tube

Weld zone

Base material

Cold-leg sintered tube
Weld zone

Base material

Attack to a depth of 16 mils

Attack to a depth of 10 mils in most areas,

complete penetration in several areas

Attack to a depth of 2 to 3 mils

Attack to a depth of 11 mils

No attack

 

Table 1.2.12. Results of Nitrogen and Oxygen Analyses of Columbium Sections

from Thermal-Convection Loop No. 58

 

Specimen

Nitrogen Content

Oxygen Content

 

(wt %) (wt %)
Arc-cast celumbium tubing insert
As received 0.009 0.045
After exposure to lithium in hot leg ot 1600°F 0.023 0,054
Sintered columbium tubing
As received 0.016 0.027
After exposure to lithium in hot leg ot 1600°F
WGId zone 0-025 0-04]
- Base material 0.027 0.035
After exposure to lithium in cold leg ot 1300°F
Weld zone 0.023 0.039
Base material 0.023 0.038

 

27

 
 

8¢

 

 

 

 

 

  

 

 

   

 

 

i 4
i /1. SINTERED _
210 - /1" TUBING
P
b 7 i
| ! 7 LITHIUM
|| FLOW
[ 85% ze-15% Cb /AN 7
[ ALLOY —*T‘—“ ! ,’/f
! | . ARC CAST
] 77 TUBING
\1 P
] 85% Zr-15% Cb
ALLOY
HOT LEG —=
€
o™ \
o™

| COLD LEG

 

Fig. 1.2.18. Columbium-Lithivm Thermal-Convection Loop Mo. 58 Which Operoted for 36 hr with the Hot Leg ot 1600°F and the Cold Leg at 1300°F.

 

     
    
 
 

 

 

e
ORNL-LR-DWG 31069

FO.002 in-

DIAMOND PYRAMID
HARDMESS

BEFORE TEST
DPH-85

 

EMLARGED VIEW OF
SINTERED TUBE WALL

SINTERED COLUMEBIUM
TUBE WALL

w73

s
i

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by ®

A

-
Nt
.i
-

BT T7 B84

o

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DIAMOND PYRAMID
HARDNESS

BEFORE TEST
DPH-10T

 

ENLARGED VIEW OF
ARC-CAST TUBE WALL

ARC-CAST COLUMBIUM
TUBE WALL

All photomicrographs show the specimens in the as-polished, unetched condition. Reduced 51.5%.

L¥0d3d S5342038d LOIrodd dNV

 
 

i
ORNL- LR-DWG 3106A E

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=

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|
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f AL TUBING |
'
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7a [H V. LITHIUM B
/f (W1 v FLOW e
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ALLOY £ ._ 1
Mt /A MRC CAST {

TUBING

 

SINTERED COLUMBIUM

£1_85% Zr-15% Cb_
: ALLOY 7

 

 

 

 

. L TUBE WALL
HOT LEG —= ! V4
- | '
od | !
od L 1444 (: -+
f—————0.04 in.————=]
~— COLD LEG ) [

 

 

 

 

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WELD JOINT FOLLOWING \\
EXPOSURE TO LITHIUM \\

F s e

f'«. '

 

 

 

SECTION A-A THROUGH SADDLE WELD AREA SHOWN ABOVE ARC-CAST COLUMBIUM
TUBE WaLL

Fig. 1.2.19. Columbium-Lithium Thermal-Convection Loop No. 59 Which Operated for 115 hr with the Hot Leg ot 1500°F and the Caold Leg at 1300°F,
All photomicrographs show the specimens in the as-polished, unetched condition. Reduced 56%.

b}
m
0
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ANP PROJECT PROGRESS REPORT

—

ARGON

41

 

 

 

 

 

 

 

 

 

 

 

 

 

3-in. —dio STAINLESS
STEEL CAPSULE

ORNL—LR—DWG 30445

-

VACUUM

| |
1!
I

| STAINLESS STEEL |
' VACUUM CHAMBER 1
I (2X4 X 4ft) '

 

 

 

 

 

 

 

 

 

 

L
| vacuum I j/VACUUM
— I | — '-——/ !
- | - I
p=— | — |
- nlintl
! |
| I
| |
| |
| |
|| 1 STAINLESS STEEL | /STArNLEss STEEL
| |
| -] |
' |
FURNACE MUFFLE —a 77 ? FURNACE MUFFLE —=| 77 77 |
I 1
| | | | | |
T T : T T |
H H ) H H !
| ] : | | |
u U I U U |
M M : M M :
i /‘ | /l |
| j |
| / 1
— .
| | |
A L j
ARGON

COLUMBIUM TUBES
(a)

Fig. 1.2.20.

the Test Capsule on Cotrasion.

at least 25 mils of intergranular penetration, and,
in several cases, there was complete intergranular
penetration in the longitudinal weld zone and in
the heat-affected zone of the tube wall. The
base material of each of the four tube walls in

regions away from the weld zone showed no
attack, ¥ The nature of the attack in the weld
and in the heat-affected zones of the tube wall

is illustrated in Fig. 1.2.21. As may be seen,
the tube wall was completely penetrated. A
grain-boundary phase seems to be associated
with the attack along the grain boundaries. As
yet this phase has not been identified; it may
be either columbium oxide or columbium nitride.
The grain-boundary phase is very difficult to
retain during normal metallographic polishing,

30

COLUMBIUM TUBES

 

()

Columbium-Lithium Corrosion Test Assembly for Studying the Effect of the Environment Around

and therefore special polishing techniques were
used on these specimens.

METALLURGICAL STUDIES OF LEAD-LITHIUM
ALLOYS

D. H. Jansen

Experimental work on a lead-lithium alloy con-
taining 0.69 wt % lithium, which is a potential
shielding material, was completed. Details of
metallurgical studies conducted on the alloy have
been reported.? An experimental lead-lithium
shield measuring 36 x 36 x 4 in. was prepared
for shielding studies at the ORNL Bulk Shielding
Facility.

 

%E. E. Hoffman et al., Lead-Lithium Shielding Alloy —
Metallurgical Studies, ORNL-2404 {in press).

 
PERIOD ENDING MARCH 31, 1958

e ", :
o e
-
HEAT-AFFECTED ZONE |~
¥ -
-~
-
“®
Zh 3
r\_\*\fl% =
\)5\0“ \
‘\
N

0.035in. {¢) OUTSIDE OF TUBE

 

{b) INSIDE OF TUBE

   
 
   
   

  

{a) TUBE WALL AT WELD-HEAT AFFECTED ZONE INTERFACE

UMCL ASSIFIED UNCLASSIFIED
Y.25448 — : Y-25447
J=—0.003 in. ——w=] f=—0.003 in.

=

  

Fig. 1.2.21. Columbium Tube Wall Following Exposure to Static Lithium for 100 hr ot 1500°F. (a) Unetched,
(&) and (c) etched with HF-HNOS-HESOJ-HZD. Reduced 41%. (Secret with caption)

31
ANP PROJECT PROGRESS REPORT

1.3. WELDING AND BRAZING

P. Patriarca

G. M. Slaughter

R. L. Heestand
E. A. Franco-Ferreira

Metallurgy Division

OXIDATION OF COAST METALS BRAZING
ALLOYS 52 AND 53

G. M. Slaughter

The ductility of o brozed joint is greatly
dependent upon its microstructure, that is, the
type and distribution of intermetallics, eutectics,
and primary solid solutions. The redistribution or
removal of the microconstituents should therefore
result in a change in ductility of the brazing
alloy. Gross depletion of intermetallics from a
brazing alloy fillet should provide a significant
improvement of its ductility.

The formation of a nickel-rich solid solution
layer on the fillet surfaces of joints brazed with
Coast Metals brazing alloys 52 and 53 during
oxidation has been reported.! Room- and elevated-
temperature bend tests on Inconel samples over-
laid with approximately 0.020 in. of brazing alloy
have indicated that this depleted layer possesses
considerable ductility and may actually provide
a cushion to absorb imposed stresses. This
condition is illustrated in Fig. 1.3.1, in which
a crack in the braze matrix may be seen to
terminate at the solid-solution interface. Pre-
liminary hardness measurements have indicated
that the hardness of the depleted zone is approxi-
mately 140 VHN as compared with an interior
hardness of 700 VHN. Thus it appears that the
ductility of the oxidized joints should improve
during service. This condition should be beneficial
in most applications.

A further study was conducted to determine the
rate of depletion of microconstituents during high-
temperature oxidation. Ten-gram cast buttons of
the Ni-Si-B alloy (Coast Metals alloy 52) and
Ni-Cr-Si-B alloy (Coast Metals alloy 53) were
made and oxidized ot 1100, 1300, 1500, 1600,
and 1700°F for 24-, 50-, 100-, 500-, and 1000-hr
periods. The depth of the depleted layer was
determined metallographically for each sample.

 

1P. Patriarca, A. E. Goldman, and G. M. Slaughter,
Afillfs Quar. Prog. Rep. March 10, 1956, ORNL-2061,
p "

32

The weight-change and depletion-depth data ob-
tained for alloy 52 are presented in Table 1.3.1,
and the data for alloy 53 are given in Table 1.3.2.
Plots of the data are shown in Figs. 1.3.2 and
133

It may be seen that the extent of depletion is
significantly greater in alloy 52 than in alloy 53.
Also, it may be seen that the depletion in alloy 52
would still continue after 1000 hr, while it appears
that the depletion has leveled off in the case
of alloy 53. However, in all cases the depth of
depletion increased with increasing temperature.
Hardness studies and chemical and spectrographic

   
  

UNCLASSIFIED
PHOTO Y-25177

 

 

 

 

o
m—|
T
O
£
A 3
. At . 0.02
o ’ WP E;-;‘.-,_' * “f oy v
i & - -. '.l '_;E
— . -t
- . -” . =
- 0.03
° s >
. - Q-
* Q
. 2
Fige 1.3.1. Broze Matrix Showing Crack Which
Terminates ot Solid-Solution Interface. As polished.
100X.

 

 
analyses in the different zones will be conducted
in an effort to more clearly understand the nature
of the oxidation.

UNCLASSIFIED

ORNL-LR-DWG 30446
0.030

_0.025

)

1700°F

Q
Q
]
Q

0.015

DEPLETION DEPTH {in

1500°F

0.005 1300°F
100°F
1400°F

 

0 100 200 300 400 500 600

TIME (hr)

700 800 9S00 1000

Fige 1.3.2. Results of Oxidation Tests of Coast
Metals Brazing Alloy 52.

I

PERIOD ENDING MARCH 31, 1958

DEVELOPMENT OF BRAZING ALLOYS
FOR LITHIUM SERVICE

G. M. Slaughter

Refractory-metal-base brazing alloys are being
investigated for a high-temperature reactor ap-
plication in which they would be exposed to
the lithium coolant. Nickel-base brazing alloys
have been shown to have poor corrosion resistance
to lithium,

A literature survey was conducted to determine
whether columbium and zirconium binary systems
possessed eutectics or minimums. The results
of the survey are summarized in Tables 1.3.3
and 1.3.4. A similar survey of the molybdenum,
titanium, and tantalum systems will be conducted,
and promising binary and ternary systems will
be investigated in corrosion and flowability tests.
It is expected that vacuum brazing will be required
to obtain suitable flow of many of these alloys,
and therefore a vacuum furnace system is being
constructed.

Table 1.3.1. Results of Oxidation Tests of Coast Metals Brazing Alloy 52

 

 

 

Weight Depletion Weight Depletion
Temperature Time Temperature Time
CF) (he) Change Dt.epth ©F) (hr) Change Depth
(%) (in.) (%) (in.)
1100 24 +0.08 <0.001 1500 24 0.00 <0.001
50 ~0.12 <0.001 50 —0.02 0.001-0.002
100 +0.05 <0.001 100 —-0.04 0.002-0.003
500 +0.07 <0.001 500 +0.16 0.004-0.006
1000 +0.03 <0.001 1000 -0.22 0.007-0.009
1300 24 +0.09 <0.001 1600 24 +0.24 <0.001
50 +0.02 <0.001 ' 50 —-0.05 0.002-0.003
100 +0.11 <0.001 100 —~0.10 0.003-0.004
500 +0.18 0.001-0.002 500 —0.28 0.007-0.008
1000 +0.12 0.001-0.002 1000 —0.35 0.014-0.015
1400 24 —0.06 <0.001 1700 24 0.00 0.002-0.003
50 +0.10 <0.001 50 -0.15 0.003--0.005
100 +0.25 <0.001 100 —0.30 0.007-0.008
500 +0.02 0.001-0.002 500 ~1.00 0.018-0.020
1000 +0.14 0.002-0.003 1000 ~1.22 0.025--0.027

 

 

33

 
ANP PROJECT PROGRESS REPORT

Table 1.3.2. Results of Oxidation Tests of Coast Metals Brazing Alloy 53

 

 

Temperature (°F) Time (hr) Weight Change (%) Depletion Depth (in.)

1100 24 +0.003 <0.001
50 +0.01 <0.001
100 +0.06 <0.001
500 +0.03 <0.001
1000 +0.18 <0.001
1300 24 +0.04 <0.001
50 +0.06 <0.001
100 +0.08 <0.001
500 +0.06 <0.001
1000 +0.10 <0.001
1400 24 +0.05 <0.001
50 +0.05 <0.001
100 +0.09 <0.001
500 +0.22 <0.001

1000 +0.22 0.000-0.002 (intermittent)
1500 24 +0.05 <0.001
50 +0.16 <0.001

100 +0.08 0.001 (intermittent)

1000 +0.006 0.003-0.004 (intermittent)

5000 +0.30 0.004-0.005 (intermittent)
1600 24 +0.15 <0.001

50 +0.08 0.001 (intermittent)

100 +0.17 <0.001

500 ~0.14 0.006~0.007 (intermittent)

1000 -0.12 0.005-0.007 (intermittent)

1700 24 ~0.007 0.002-0.003 (intermittent)
50 ~0.02 <0.001

100 +0.14 0.002-0.004 {intermittent)

1000 —-0.32 0.008-0.010 (intermittent)

5000 ~0.41 0.008-0.010 (intermittent)

 

34

 

 
 

PERIOD ENDING MARCH 31, 1958

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

The Zr-Cb and Zr-Cb-Mo alloys have been used Table 1.3.4. Zirconium Binary Systems of Potential
as filler materials in the welding of columbium Interest as Brazing Alloys
thermal-convection loops, and both alloys have
shown excellent resistance to corrosion by lithium Eutectics or Minimums*
(see Chap. 1.2, this report). Further studies of System Composition Melting Temperature
these alloys will be conducted, and the Zr-Mo (wt %) CF)
and Zr-Ti alloys, which should also be corrosion
resistant, will receive early attention. Zr 3353

Zr-Al 11 Al, eutectic - 2460
UNCLASSIFIED 22 Al, eutectic 2700
ORNL -LR-DWG 30447

0020 ! 46 Al, eutectic 2700
£ oots - | R | Zr-B 2 B, eutectic 3200
E | . 38 B, eutectic 4208
S ' f700°F 75 B, eutecti 3380
g 0.010 |- 1 . /L____.—T —— 1 o y eutectic 3
F // 1600°F Zr-Be 5 Be, eutectic 1800
& 0005 |- 1 — o HB00°F - | oo e, eutectic 5
2 =" [1s00°F_ 93 Be, 2250
a . """ 1300°F

o ok " | KES:)D'F HOOTF Ty, 1 Zr-Cb 25 Cb, minimum 3200
0 100 200 300 400 500 600 700 80O 900 {0CO Zr-Co 20 CO, eutectic 1750
TIME (re) 45 Co, eutectic 2750

65 Co, eutectic 2700

Fige 1.3.3. Results of Oxidation Tests of Coast Ze-Cr 18 Cr, eutectic 2340

Metals Brazing Alloy 53. 69 Cr, eutectic 2970
Zr-Cu 21 Cu, eutectic 1830
37 Cu, eutectic 1710
Table 1.3,3, Columbium Binary Systems of Potential 47 Cu, eutectic 1635
Interest as Brazing Alloys Zr-Fe 16 Fe, eutectic 1715
84 Fe, eutectic 2426
Eutectics or Minimums* )
Zr-Ge 7.7 Ge, eutectic 2795
System . Melting 99 Ge, eutectic 1712
Composition Temperature ‘

(wt %) CF) Zr-In 22 In, eutectic ~ 2550
Zr-Mn 22.5 Mn, eutectic 2076
Cb 4475 Zr-Mo 31 Mo, eutectic 2768
Cb-Co 79 Co, eutectic 2300 Zr-Ni 17 Ni, eutectic 1760
27 Ni, eutectic 1805

Cb-Fe 33 Fe, eutectic 2850
82 Fe, eutectic 2500 Zr-5b 22 5b, eutectic 2600
Cb'Ni 48 NI, eutectic 2]50 ZI"Si 2-4 Si, eU"eCfiC 2940
77 Ni, evtectic 2320 75 Si, eutectic 2470
Cb-Si 18 Si, eutectic 3450 Zr-Sn 23.5 Sn, eutectic 2894
58 Si, eutectic 3400 Zve-Th 75 Th, minimum 2426
95 Si, eutectic 2330 Zr-Ti 55 Ti, minimum 2900

Cb-Th 93 Th, eutectic 2600 Zr-V 30 V, eutectic 2246

Cb-v 65 V, eutectic 3290 Zr-W 18 W, eutectic 3000

Cb-Zr 75 Zr, minimum 3200 Zr-Zn 31 Zn, eutectic 1860

*J.P. Page, Alloys and Compounds of Niobium, ORNL *B, Lustman ond F. Kerze, Jr. (eds.), The Metallurgy
CF-56-10-36 {Oct. 8, 1956). of Zirconium, 1st ed., McGraw-Hill, New York, 1955,

35

 

 
ANP PROJECT PROGRESS REPORT

1.4. MECHANICAL PROPERTIES

D. A. Douglas

Metallurgy Division

STRAIN-CYCLING INVESTIGATIONS

Investigations of the problem of thermal fatigue
in reactor materials were continued. The capacity
of Inconel to absorb strain reversals is being
intensively studied in the temperature range from
1200 to 1650°F.
as strain amplitude, temperature, cycle frequency,
grain size, geometry, and environment are being
considered. Three types of test have been em-
ployed in these investigations: mechanical strain
cycling, thermal strain cycling, and mechanical
stress cycling (fatigue). The mechanical-strain-
cycling tests are being conducted at ORNL. The
other two test programs have been subcontracted
to The University of Alabama (thermal strain
cycling) and Battelle Memorial Institute (fatigue).

Much of the information obtained thus far from
mechanical-strain-cycling tests has been reported

The effects of such variables

in previous reports in this series.'=3 The con-
clusions that have been drawn from the test
results are (1) that the relation between total
plastic strain absorbed per cycle (€.} and the
number of cycles to failure (N} is of the form

a —
(n NeP—C,

where a and C are constants; (2} that grain size
is a significant factor, with fine-grained material
being superior to coarse-grained material in its
ability to survive a greater number of cycles; and
(3) that temperature does not have a large effect
on time to rupture in the range of 1300 to 1600°F
at cycle frequencies of 0.5 per minute.

Recent investigations have been concerned with
the effect of the cycle time (frequency) on the
relationship between € _and N. In these tests a
load is applied which is of a magnitude that will
cause the specimen to experience a specified
plastic strain in one half of the specified cycle

 

‘C. R. Kennedy and D. A. Douglas, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 190.

2c. R. Kennedy and D. A. Douglas, ANP Quar. Prog.
Rep. Sept. 30, 1957, ORNL-2387, p 182.

3D. A. Douglas et al., ANP Quar. Prog. Rep. Dec. 31,
1957, ORNL-2440, p 153,

36

R. W. Swindeman

time, and then the load is reversed to produce -
the same plastic strain in the opposite direction
in the other half of the cycle time. Comparisons
of data obtained at different frequencies and
temperatures are presented in Figs. 1.4.1 through
1.4.4, As may be seen in Fig. 1.4.1, the results
for specimens tested in argon at 1300°F with
2- ond 30-min cycles give €, vs N curves that
fall close together. When the plastic strain per
UNCLASSIFIED

DRNL -LLR-DWG 30448
100

o
Q

¢ 2-min CYCLES
o 30-min CYCLES

N
o

S

i8]

€, PLASTIC STRAIN PER CYCLE (%)
Qo
Mmoo~ N

o
o

 

o

01 02 05 ¢ 2 5 1o 20 50 100 200 500 1000
N, CYCLES TO RUPTURE

Fig. 1.4.1. Effect of Strain Cycling on Fine-Grained
Inconel Tubes Tested in Argon at 1300°F.

UNCLASSFIED
ORNL-LR-DWG 30419

100
>0 2 min CYCLES
{0-min CYCLES
20 20-min CYCLES

30-min CYCLES

05

¢,, PLASTIC STRAIN PER CYCLE (%)

02

 

Oa
2 5 10 20 50 100 200 500 1000 10,000

N, CYCLES TO RUPTURE

Figl ]-402.
Inconel Tubes Tested in Argon at 1500°F.

Effect of Strain Cycling on Fine-Grained

 

 
UNCLASSIFIED
ORNL-LR-DWG 30420

8

S
o

® 2-min CYCLES
4 {0~ min CYCLES
4 20-min CYCLES
© 30-min CYCLES

n
Qo

o

[6)]

PLASTIC STRAIN PER CYCLE (%)
O
o LN

v,

o
o

 

o

1 2 5 10 20 50 100 200 500 1000

N, CYCLES TO RUPTURE

10,000

Fige 1:4.3. Effect of Strain Cycling on Coarse«Grained
Inconel Tubes Tested in Argon at 1500° F.

UNCLASSIFIED
ORNL -LR-DWG 30421
100

50

—- n
O O

w\

- n

€p, PLASTIC STRAIN PER CYCLE (%)
o
o

0.2

 

b2 5 40 20 50
N, CYCLES TO RUPTURE

100 200 500 1000

Fig. 1.4.4. Effect of Strain Cycling on Fine-Grained
Inconel Tubes Tested at 1500, 1650, and 1900°F in

Argon at 30 Minutes per Cycle.

cycle is above 1%, the 30-min cycle results in a
slight increase in the number of cycles before
failure. The effect of long cycle periods at low
strains (below 1%) remains to be clarified.

In similar tests of coarse- and fine-grained
specimens at 1500°F, cycling periods of 2, 10,
20, and 30 min were used. The results of these
tests, as presented in Figs. 1.4.2 and 1.4.3, show,
in all cases, that when the strain per cycle is

 

PERIOD ENDING MARCH 31, 1958

above 5% the cycle frequency has no significant
effect on the number of cycles to rupture. At low
valves of €, however, the cycle time becomes
increasingly important, At the same plastic strain
per cycle, the number of cycles to rupture de-
creases with increases in the cycle period. From
a standpoint of actual time to rupture, however,
long cycle periods result in increased life. Since
the 30-min cycle period is 15 times greater than
the 2-min cycle period, a considerable reduction
in the number of cycles to failure would have to
occur before the time to rupture would be shorter
for the 30-min cycle than for the 2-min cycle.

A comparison of results obtained at 1300, 1500,
and 1650°F with a 30-min cycle is given in
Fig. 1.4.4. The data indicate that the ‘'break
point,”’ that is, the point where a frequency effect
is first noticeable, moves to higher values of ¢
as the test temperature is increased. These data
contrast with data for 2-min cycles, which showed
little effect of temperature. The existence of a
break in the € vs N curves is somewhat ques-
tionable, since it is possible that a smooth curve
which did not satisfy the empirical relation given
by Eq. 1 could be drawn through the data.

Actually, in the type of test system employed
in these frequency investigations, the value of
€, is not constant throughout the test; a small
increase in €_occurs as the test progresses. In
general, the change in € occurs in the first few
cycles and thereafter increases only slightly.
The values of € which are given in the data
correspond to measurements taken in the middle
or near the end of the test, and therefore the
change in €, that occurs in the early part of the
test is not reflected to a |arge extent in the curves
presented in Figs. 1.4.1 through 1.4.4.

Since the cyclic strains of greatest interest from
a design standpoint are well below 5%, it appears
that cycle frequency must be considered in
evaluating the probable lifetime of a material that
will be subjected to cyclic strains. Further work
is needed therefore to establish the effects of
strain cycling over a larger frequency spectrum
and a larger strain amplitude range.

Since the cyclic strains induced during the
operation of a reactor are primarily the result of
thermal rather than mechanical strain cycling, the
mechanical-strain-cycling data can be useful in
design only if correlation can be obtained with
results from comparable thermal-cycling tests.
Tests in which the strain is induced thermally

37

 
ANP PROJECT PROGRESS REPORT

have been conducted at mean temperatures of 1300
and 1500°F with a cycle period of 38 sec. Data
obtained at 1300°F in the two types of tests show
excellent agreement when analyzed on the basis
of the equation relating plastic strain to the
number of cycles to failure.? Data obtained in
similar tests at 1500°F are compared in Fig. 1.4.5.

UNCLASSIFIED
ORNL-LR-DWG 30422

o

. MECHANICAVL- STRAIN - CYCLING DATA
o THERMAL - STRAIN - CYCLING DATA

1

ey, PLASTIC STRAIN PER CYCLE (%)

 

10 20 50 100 200 500 1000 10,000 100,000
N, CYCLES TO RUPTURE

Figs 1.4.5. Comparison of Effects of Thermal ond
Mechanical Strain Cycling on Inconel Tested at a Mean

Temperature of 1500° F.

As may be seen, the correlation of the data from
the two tests is poor, especially at values of ¢
greater than 0.1%. The lack of agreement may
be attributed, in part, to difficulties in determining
the plastic strain in fthe thermal-strain-cycling
test. The specimens are resistance heated in
the thermal-cycling tests, and high thermal gra-
dients are set up that produce variations in the
elastic and plastic properties. Therefore an
‘‘effective gage length’' must be established
which may vary as the test progresses. Local hot
spots produced by imperfections, cracks, or
changes in cross-sectional area also produce
localized strain. These effects are most pro-
nounced for the high-temperature cycles, which
correspond to the high-strain cycles. More refined
methods of determining the plastic strain and of
minimizing the temperature gradients are being
considered.

Some data have also been obtained concerning
the effect of the cycle frequency on the life of
material subjected to thermal strain cycling. Two
tests have been conducted at 1300°F for which
30-min cycles were used. One specimen was

cycled at 1.3% €pr and it survived 100 cycles

38

as compared with 600 cycles in a test for which
the cycle period was 38 sec. The second
specimen, which was cycled at 0.93% € , lasted
375 cycles as compared with 800 cycfes when
tested on a 38-sec cycle.

In some instances, service conditions may exist
within a reactor in which the cyclic frequencies
of the strains, whether mechanically or thermally
induced, will be high, that is, 1 to 10 eps. High-
temperature fatigue may occur under such con-
ditions that is usually considered to be the result
of stress cycling rather than strain cycling.
Fatigue studies are therefore being conducted on
Inconel specimens at 1200, 1400, and 1600°F.
Results from tests at 1600°F in which frequencies
of 1 and 10 cps were used are shown in Fig. 1.4.6.
Under these conditions, the effect of cycle
frequency is evident. Specimens cycled at 10 cps
endure 10 times as many cycles at the same
stress level as those cycled at 1 cps. An attempt
is being made to dynamically measure the strain
per cycle at these frequencies. If this can be
done, the results of fatigue investigations can
be correlated with the strain-cycling data.

UNCLASSIFIED

ORNL-LR-DWG 30423
16,000

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

= | ]
® | ‘
L . ¢ ! | :
n 14,000 L N | e .
& \c.;tcps ! \<J|Ocps ! .
&
& 12,000 N S
o 1 I~ \L
Z M~ | T
S T — | |
£ 10,000 {—re =
) i
5 ‘ L
< i :

8000 f i

w2 5 0% 2 5 10® 2 5 0o

#, CYCLES TO RUPTURE

Fig. 1.4.6. Effect of Stress Cycling on Inconel
Tested at 1600° F.

One particular question which can be answered
by fatigue studies is whether high-frequency strain
cycling weakens the material being tested. At
frequencies such as those employed in the strain-
cycling tests (0.5 cpm), weakening of the Inconel
specimens was observed at and above 1300°F,
and the weakening progressed as the number of
cycles increased. In high-frequency fatigue tests,
however, the strain rate is very high, the recovery
time is minimized, and the conditions are favorable
for strain hardening. There is evidence that small
cyclic stresses strengthen Inconel from a creep
standpoint, as may be seen in Fig. 1.4.7. It also
may be seen, in comparison with the results
UNCLASSIFIED
ORNL-LR-DWG 30424

100

50

20

STRAIN (%)
M o O

05

c2

oA

 

TIME (hr)

Fig. 1.4.7.
Tested at 1600°F in Air at 4000-psi Mean Stress and

Several Stresses Alternated at 1 eps.

Creep Curves of Coarse-Grained Inconel

obtained for a mean stress of 4000 psi, that
application of a stress of 1000 psi alternated at
a frequency of 1 cps resulted in a decrease in
creep rate and an increase in time to rupture.
Higher stresses, 2800 and 8000 psi alternated at
1 eps, however, decreased the time to rupture,
On the basis of these and similar tests at other
mean stress levels, it appears that low alternating
stresses are beneficial with regard to creep
characteristics.

Goodman diagrams® are presented in Figs. 1.4.8
and 1.4.9 that are based on the results obtained
in the tests at 1 and 10 cps, respectively. In
these diagrams, selected times to rupture are
plotted as a function of the ratio of the alternating
stress to the mean stress. From these plots the
stress conditions that will result in failure by
creep can be differentiated from the stress con-
ditions that will result in failure from fatigue in
a system in which both creep and fatigue exist.

CREEP UNDER BENDING STRESSES

Studies have been initiated for determining the
extent to which tensile creep data can be cor-
related with data on creep under bending stresses.
The testing apparatus shown in Fig. 1.4.10 has
been designed to impose a pure bending stress

 

45. Timoshenko and G. H. MacCullough, Elements of
Strength of Materials, 3d ed., p 371, Van Nostrand,
New §ork, 1949.

PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x 107 ORNL—-LR— DWG 30425
14
A=RATIO OF THE ALTERNATING
STRESS TO THE MEAN STRESS
t = TIME TO FAILURE
STRESS ALTERNATED AT 1cps
|
O hr A4=2.0
<f = {00 hr
- \ f =10 hr
w
s \
w
- X "/
Z / \ V
i © C 7 ' /7)‘/ '
=
o
w
2 / \
4 fi/ t =200 hr- p\
~02%
/ \ e
2 S S -
/
yd s
/’/
O O
0 2 a4 6 8 10 (x103)
MEAN STRESS (psi)
Figs 1.4.8. Goodman Diagram (Time to Failure) for

Inconel Tested at 1600° F.

Stress alternated at 1 cps.

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x 10%) ORNL— LR— DWG 30426
44 T T T
A= RATIO OF THE ALTERNATING
STRESS TO THE MEAN STRESS
t = TIME TO FAILURE
12 f=10hr =~ STRESS ALTERNATED AT 1 cps
f =50 hr A-_-Z.O
— 10 NG
»
o
0 b
'&J g8 |- N ,,,/
j—
@ =100 hr-f-/X\ 1 v
0 o
Z z
= p
g 66— — -1 A4 N -
Z / \ r
o
wJ
—
|
< a t=200hr b(
_02%
2 At o |
/ // }:
0
0 2 4 6 8 10 {x 103)
MEAN STRESS (psi}
Fig: 1.4.9. Goodman Diagram (Time to Failure) for

Inconel Tested at 1600° F. Stress alternated at 10 cps.

39

 
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
PHOTO—42574

     

BEAM SPECIMEN

 

Figt 1.4.10.
Bending Stresses.

Apparatus for Investigating Creep Under

on a beam specimen. A similar aopparatus was
used by Platus and Cotton® for room-temperature
tests of lead. The beam specimen is a 0.20-in.-
square bar of Inconel with a l-in. gage length.
It fits tightly into square notches cut into the

 

SD. L. Platus and B. Y. Cotton, Prediction of Creep
Failures of Lead in Bending from Tensile Data, ORNL
CF-57-7-58 (July 17, 1957).

40

small disks shown in Fig. 1.4.10. Couples that
tend to rotate the disks in opposite directions
are applied by means of equal loads on the reeds.
The couples produce a bending moment in the
beam specimen. The stress in the outer fibers
of the beam, o, as calculated from elastic theory
is
3DP

o= ——=75P 5

w3

where D is the diameter of the specimen-holding
disk (2 in.), P is the applied load, and w is the
thickness of the beam specimen.

The tests have been conducted in fused-salt
environments in order to obtain uniform temper-
atures in the test sections. As bending occurs,
the disks rotate, the reeds unwind, and the pull
rods move upward., The beam deforms as an arc
of a circlee A simple relation exists in this
case between strain in the outer fiber, £, and the
upward travel of the pull rod, x:

wx x

2 10
where w is the beam width (0.20 in.) and / is the
gage length (1 in.).

The creep-in-bending curve for Inconel tested
in a fused-salt environment at 1500°F and stressed
at 6000 psi is presented in Fig. 1.4.11, and the
tensile creep curve at the same stress level is
presented for comparison. For the first 10 hr,
the agreement is satisfactory between the results
of the two tests. After this period, however, the

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
25 ORNL-LR-DWG 30427
20
: e L
= 45 " =
z 7
<
£ &7 sl
2 40 @ 5
.5}\
& P?
& o
¢ i
5 e
{Q}f}w !
T
o ‘
0 TE.I

 

 

 

 

 

 

 

0 50 100 1S 200 250 300 350 400
TIME (hr)

Fig. 104111-
Fine-Grained
vironment at 1500° F.

Tensile and Bending Creep Curves for

Inconel Tested in a Molten-Salt En-

 
slope of the tensile creep curve begins to increase
rapidly. According to Platus,® the bending stress
is very quickly redistributed in the beam in the
early stages of creep, and there is a resulting
decrease in the outer fiber stress to about 80%
of the value predicted by elastic theory. |f this
analysis is correct, the bending creep rate after
a few hours should be comparable with the tensile
creep rate at a stress of 4800 psi. As may be
seen in Fig. 1.4.11, the agreement is good, to a
first approximation, for strains up to 10%. The
slopes of the tensile creep curves can be expected
to increase more rapidly, however, than the slope
of the bending creep curve, since corrosion by
the fused salt, cracks, and changes in cross-
sectional area play greater roles in changing the
creep rate of the 0.060-in.-thick tensile sheet
specimens than in changing that of the 0.20-in.-
thick bending creep specimen.

CREEP UNDER MULTIAXIAL STRESSES

Uniaxial creep tests have yielded valuable data
on the creep characteristics of structural materials
at high temperatures, but such information is not
applicable under service conditions which produce
multioxial stresses in many reactor components.
The creep and rupture properties of materials
under such multiaxial stress conditions are there-
fore of interest, especially where the data can
be used to evaluate various mechanical theories
of flow and fracture. Creep tests on multiaxially
stressed Inconel tubing are being conducted at
1500°F in which the principal stresses, that is,
the axial and tangential stresses in the tube wall,
are varied by imposing an axial load on the
pressurized tube.” The radial stresses are small
as compared with the axial and tangential stresses
and are therefore neglected; the remaining axial
and tangential stresses are thus referred to as
biaxial rather than triaxial. The stress distri-
bution in the tube wall is described by the ratio
of the axial stress to the tangential stress, crz/crt.
The time to rupture and the creep rate are ob-
served as the stress ratio is varied from specimen
to specimen.

 

p. L. Platus, ANP Quar. Prog. Rep. March 31, 1957,
ORNL-2274, p 6.

7C. R. Kennedy and D. A. Douglas, ANP Quar. Prog.
Rep. Sept. 30, 1957, ORNL-2387, p 185.

PERIOD ENDING MARCH 31, 1958

Tests are being conducted in which one of the
principal stresses is held at 3000, 4000, or
6000 psi while the stress ratio is varied from
o to —1.0. Axial creep curves obtained in the
tests at 6000 psi are presented in Figs. 1.4.12
and 1.4.13. The data show a trend similar to that
found previously in tests at 4000 psi.” As long
as the axial stress is held at 6000 psi (Fig.
1.4.12), the slopes of the creep curves are prac-
tically independent of the tangential stress. The
time to rupture and the strain at rupture do,
however, show some dependence on the stress
ratio. Both the time to rupture and the strain at
rupture decrease as the stress ratio approaches
1.0. For ratios below 1.0, the axial stress is
decreased, and the axial creep rate is, therefore,
correspondingly decreased. The axial strain at
rupture and the time to rupture continue to de-
crease, as shown in Figs. 1.4.14 and 1.4.15. At
a stress ratio of 0.5, no appreciable creep in the
axial direction can be detected. For stress ratios
lower than 0.5, compressive creep occurs, as
shown in Fig. 1.4.13, The strain in compression
at rupture increases with decreasing axial stress.

UNCLASSIFIED
30 ORNL—LR—DWG 30428
20 o /O'f = 6000/0

6000/1000
6000/ 3000—,

6000/6000

5000/6000

AXIAL STRAIN (%)
o

-

4000/6000

0.5

02
3000,/6000

 

0.4
10 20 50 100 200 500 1000
TIME (hr)

Figs. 1.4.12. Creep Curves of Inconel Tubes Biaxially
Stressed and Tested in Argon at 1500° F.

4]

 
ANP PROJECT PROGRESS REPORT

Any theory that is an attempt to explain behavior
under biaxial stresses must relate the variation
in creep rate and rupture life to the stress ratio.
As indicated previously and shown in Fig. 1.4.16,
multiaxial creep rates at stress ratios greater than
0.5 correlate well with uniaxial creep rates at the
same stress level. The data are plotted without
consideration of the tangential stress. The creep
rates were token from creep curves at 1% strain.
If the increasing creep rate in the tangential

UNCLASSIFIED
ORNL—LR—DWG 30429

 

20
8 -
z = —6000/6000
= 10 %/ /
o
wr
w
w
& 5
=
o
o
z
2
I 2
—
u
—
= 0/6000
=

0.5

1 2 5 10 20 50 100
TIME (hr)

Fige 1.4.13. Creep Curves of Inconel Tubes Biaxially
Stressed and Tested in Argon at 1500°F.

in compression.

Axial strain

UNCLASSIFIED
ORNL—LR— OWG 30434

 

32

2

 

 

 

 

 

 

 

 

¢, STRAIN (%)
O @D

./< Ne
>

 

 

 

 

 

 

 

 

 

 

-24

-6000,/6000 6000/0

0/6000 6000,/6000
STRESS RATIO, o, /o,

Fige 1.4.15. Total Strain at Rupture of Inconel Tubes
Creep Tested ot 1500°F Under Combined Axial and
Tangential Stresses,

UNCLASSIFIED
ORNL-LR-DWG 30430
5

 

 

6000/0

L

 

6000 /2000

[ &

 

6000/4000

 

6000,/6000

 

4000/6000
o0

/
/
/e.

 

STRESS RATIO, o, /o,

2000/6000 /
0/6000

 

 

- 20006000
/ .

 

 

 

-4000 /6000 /

 

 

 

 

- 6000/6000 .
0 100

200 300 400

TIME TO RUPTURE ({(hr)

Fig. 1.4.14. Time to Rupture as a Function of the Stress Ratioc in Inconel Tubes Bioxially Stressed and Creep

Tested in Argon at 1500° F.

42

 

 
direction, which corresponds to the decreasing
stress ratio, were to induce a compressive com-
ponent to the creep in the axial direction, the
observed creep rate would decrease as O’z/O’t
decreased from = to 1. Such behavior would be
predicted under the assumption that the sum of
the strains and of the strain rates in the three
principal directions must at all times be zero.®
No influence of tangential creep is observed.
Moreover, as the axial stress decreases and the
stress ratio drops from 1.0, the axial creep rates
remain in agreement with pure tensile data. No
effect of the tangential strain is noted until the
stress ratio approaches 0.5. Here, the sudden
disappearance of axial creep can be partly at-
tributed to foreshortening of the tube caused by
““bowing out” in the center of the gage length.
Below the ratio 0.5, the axial creep rate in com-
pression is higher than that which would be
expected by assuming dependence on axial stress
alone. Two additional factors influence this creep

 

8. Gensamer, Strength of Metals Under Combined
Stresses, American Society for Metals, Cleveland, 1941.

STRESS ({psi)

102
1078 2 5 1073 2 5

 

Tl 2

PERIOD ENDING MARCH 31, 1958

rate, however. One of these is the bowing-out
effect previously mentioned. The other is an
eccentric-load effect from the compressive axial
load acting on the bowed-out tube. It is difficult
to predict how creep in one direction may affect
the deformation rate in the other., For one thing,
deformation is accompanied by changes in the
stress values and stress distribution, which, in
turn, result in changes in the creep rates in the
principal directions. Evidence of stress redis-
tribution has been found during tests. Near the
stress ratio of 0.5, specimens have been observed
to start creeping in compression and then move
into tension. The opposite effect has also been

noted.

On a microscopic scale, there are many factors
which have not been considered in the theories
of flow. For example, the generation of cracks,
voids, and lattice defects by the maximum stress
may influence the creep rate produced by a lesser
stress in another direction. The stress de-
pendence of the creep mechanism could also be
considered. Preferred orientation, which is
present to a degree in tube material, causes an

UNCLASSIFIED
ORNL—LR—DWG 30432

ay fop = © (UNIAXIAL)

o‘z/o'f =2
z/% =1
05< az/crf("
c'z/crr: -0.5

—2

CREEP RATE (hr™ %)

Fig- 104-]60

Tangential Stress Ratios.

Axial Creep Rate of Inconel Tubes at 1500°F When Bioxially Stressed at Various Axial-to-

43
ANP PROJECT PROGRESS REPORT

dependence of elastic ond plastic
properties. Factors that influence the fracture
point of the material are even more numerous. At
the present time no clear fracture criterion has
been established.

orientation

The maximum shear stress,

44

the octahedral shear stress, and the maximum
normal stress theories of yielding have been
rejected because they fail to explain the type of
stress-ratio dependence which has been observed

for the time to rupture.

 
 

PERIOD ENDING MARCH 31, 1958

1.5. CERAMICS

L. M. Doney

R. L. Hamner

R. A. Potter

Metallurgy Division

In cooperation with GE-ANPD, fabrication studies
are being made of dense beryllium oxide for use
as a neutron reflector material. The possibility
of adding boron (0.5 to 1 wt %) as a suppressor
for secondary gamma rays is also being inves-
tigated. The two phases of the studies involve
(1) a survey of boron compounds and their effects
on the thermal stability and oxidation resistance
of BeO ond (2) densification of extruded and
cold-pressed BeO to which the boron compounds
have been added.

BORON COMPOUNDS

The boron compounds of initial interest were
the borides of the transition metals of the fourth,
fifth, and sixth periodic groups, whose physical
properties, such as thermal stability and oxidation
resistance, made them particularly attractive for
the application described. For preliminary studies
it was necessary to synthesize the compounds in
the laboratory. Diborides of Ti, Zr, Hf, Ta, V,
and Cr were synthesized by reacting small batches
(~ 100 g) of the metal oxides, or, in the case of
zirconium, the metal silicate, with graphite and
boron carbide in graphite crucibles heated by
high-frequency induction to temperatures in excess
of 2000°C. Because of reaction-product gases, it
was difficult to determine the exact temperatures
of the reactions by the available optical methods;
therefore, visual observation of ignited out-gases
and time-power control of the induction unit
aided in determining reaction progress.

The reaction between zircon sand, graphite,
and boron carbide was more complete than the
other reactions, and the product was cleaner.
During the reaction, ZrB2 was produced as a
mass of granules slightly sintered together which
were similar in size and shape to the original
sand particles. X-ray diffraction patterns were
made of all the compounds, and in each case
the material was found to be predominantly
boride.

BORON ADDITIONS TO BeO

Specimens of boron-containing BeO were pre-
pared for preliminary evaluation by mixing the

boron compounds with Brush Beryllium Company
“Luckey S.P."" grade BeO in sufficient amount
to give 1 wt % boron and then hot pressing the
mixtures at 1900°C at a pressure of 2000 psi.
Specimens containing ZrB,, TaBz, TiBz, CrBz,
HfBz, CeB,, and BN were prepared.

Comparable densities (94 to 96% of theoretical)
were obtained for all mixtures except CeBsoBeO,
which was discarded because of its generally
poor appearance. Microscopic examinations indi-
cated that the additives were held in the BeO
and uniformly distributed, with the exception of
CrB,, the lowest melting compound of the series,
which appeared to have migrated toward the
periphery of the specimen.

The fabricated specimens were oxidation tested
at 1000°C and at 1300°C. At 1000°C the ZrB.,,-
BeO and HfB,-BeO specimens showed no weigint
change after 34 hr; the TaB,, TiB,, and CrB,
combinations showed small weight changes
(<0.2%) and attendant oxidation products. The
exposed BN of the BN-BeO specimen had almost
completely disappeared, and there were easily
visible pits and craters. The disappearance of
the BN is believed to be due to oxidation and
subsequent volatilization.

The ZrB, and HfB,, mixtures exhibited oxidation
resistance at 1300°C that was much superior to
the oxidation resistance of other combinations.
The ZrB, ond HfB, specimens showed weight
changes of only 0.15% after 300 hr. X-ray patterns
of the ZrB,-BeO mixture before and after oxidation
were identical.

In view of its stability in air and its avail-
ability, only ZrB, will be used in future work
on this phase of the program. Specimens of the
BeO-ZrB, mixture are being fabricated for determi-
nations of modulus of rupture, compressive creep,
and thermal conductivity before tests are made

in the ETR.

DENSIFICATION OF BeO

Additives are being sought that will densify
extruded and cold-pressed BeO bodies and
possibly, but not necessarily, fulfill the boron

45

 
 

ANP PROJECT PROGRESS REPORT

content requirements. Preliminary data obtained
by the Brush Beryllium Company indicated that
the following two compositions should be investi-
gated:  94.0% BeO-5.0% MgO-1.0% B,C and
98.0% BeO-1.0% Fe,0,-1.0% B,C. An extruded
body was produced from the first composition
that, when fired to 1500°C in an oxidizing at-
mosphere, had a density of approximately
2.86 g/cm?. No determination was made of the
boron content.

In order to determine effects of the individual
components on BeO, mixtures were made con-
taining the following parts by weight: 94.0 parts
BeO plus 5.0 parts MgO; 94.0 parts BeO plus
1.0 part B,C; 98.0 parts BeO plus 1.0 part Fe,O,;
98.0 parts BeO plus 1.0 part B,C. Data obtained
here have indicated that the additive ZrB, may
possibly fulfill the density requirement, as well
as the boron requirement, and, therefore, the

46

composition 94.8% Be0-5.2% ZrB, is also being

investigated.

The various compositions, including one of
100% BeO for comparison, were mixed in a jar
on rolls for 18 hr and then pressed at 20,000 psi
in steel dies, broken up, granulated by passing
through a 30-mesh screen, and remixed for 2 hr.
Right cylinder shapes approximately ¥ x ¥ in.
were made by pressing the materials at 15,000 psi.
Pieces from each of the compositions are being
fired to 1100, 1300, and 1500°C in oxidizing
atmospheres. These pieces will be compared with
others fired in inert atmospheres and their dif-
ferences in density and boron retention properties
will be noted. In the future, these compositions,
as well as additional ones, will be made into
extruded shapes and compared with their cold-
pressed counterparts.
PERIOD ENDING MARCH 31, 1958

1.6. NONDESTRUCTIVE TESTING
J. W. Allen

R. W. McClung

R. A. Nance

Metallurgy Division

EDDY-CURRENT THICKNESS MEASUREMENTS
J. W. Allen R. A. Nance

The study of eddy-current techniques for meas-
uring cladding thicknesses on flat fuel plates
was continued, with special emphasis on the
ability of the probe coil to locate areas of
minimum cladding thickness.
graphic correlation studies,! sections have been
taken through a point beneath the center of the
probe, and the results have been erratic because
the area of thinnest cladding does not necessarily
occur at this point. In order to gain a better
understanding of this problem, a study was made
of the cladding thickness variations over the

In past metallo-

entire ]/2-in.-dia area underneath the probe coil.
This was accomplished by successively removing
thin layers of the cladding and plotting contours
of the cladding thickness wvariations. Contour
plots of two such areas which were examined
are presented in Figs. 1.6.1 and 1.6.2. The
several sections taken through the area beneath

UNCLASSIFIED
ORNL-LR-DWG 27533

PLAN VIEW

CLADDING THICKNESS
LEGEND

CORE: 48 wt% U
3 wt% Si

49 wt% Al
CLADDING: 100 Al

0.0065—-0.008n.

0.008—0.010in.

LOCATION OF PROBE COIL

0.040—0.0115 in. (COIL DIAMETER %g in.}

Ot W CLADDING THICKNESS
MEASURED BY
EDDY -CURRENTS = 0.014 in.

0.0115—-0.014 in.

   

 

   
 
 
     

  
 
  

 

 

       

 

>0.014in, INCHES
i i 3
cogs | colLoa “ ‘2 /4
: gy s CLADDING
0.010 [ F
- A T e CORE
:> 0_0’4 Y. //Il/ll/”l’l’/’/ll”’;”’.
= SECTION A—A
&
2 0006 [ LGOI DA
ST "L~ CLADDING
I o010 [=
® — CORE
Zz 0044
=
a
Ld
S 0006
o E CLADDING
0.010 —————
— BTt e s CORE
A SLT LTS ILS S ST IR LSS ITS ST S ITS S ST IS

 

Q.04
SECTIONC-C

Fig. 1.6.1. Contour Plot of Clodding Thickness

Yariations on Flat-Plate Fuel Element.

the center of the probe indicate the varied meas-
urements which can be obtained by changing the
orientation of the section. Each section indicates
a different minimum cladding thickness, none of
which occurred beneath the center of the probe.

The study of the possibility of measuring the
thickness of a hot-dipped aluminum coating on
steel was also continued. This metal system
is not of any particular project importance in
itself, but it is typical of metal systems in which
a magnetic material is clad with a nonmagnetic
material. The specimens that were examined had
aluminum coatings of 0.0016-in. nominal thick-
ness, including an intermetallic layer of approxi-
mately 0.0003-in. thickness, as shown in Fig.
1.6.3. The intermetallic layer adversely affects
the measurement of coating thicknesses of less
than 0.0005 in., but the accuracy of the measure-
ment is +0,0002 in. for the thickness range of
0.0005 in. to 0.0016 in. Measurements of aluminum
coating thicknesses on steel in the absence of

 

]R. B. Oliver, R. A. Nance, and J. W. Allen, ANP
Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 192.

UNCLASSIFIED
ORNL-LR-DWG 27534

PLAN VIEW

CLADDING THICKNESS
LEGEND

CORE: 49wt % U
3wt % Si
49 wt % Al
CLADDING: 1100 Al
c

0.0135—0.015in. ¢

LOCATION OF PROBE COIL

0.015 — 0,047 in.
" {COIL DIAMETER %g in.)

CLADDING THICKNESS

 

 

 
   
  

 
     
   

  
  

 
  
 

 

  

  

>0.047 in.
MEASURED BY
EDDY-CURRENTS = 0.044 in.
INCHES
0 Ya Ya Ya
COIL DIA ———=q
0.013 T CLADDING
- T Y
£ o.oie B e CORE
w SECTION A—A
@
W
z jw———COIL DIA——=
s 0.043 RS ———_==- CLADDING
I _.Il-l'l;};;l’l’;;;}lllltl’lfl: CORE
— O X 04 e O e e B L L Ll L LD
o SECTION B-B
=2
2
o —— COIL DIA——=
< 0.013 HT; T CLADDING
© o e T
0.01g BRI CORE
SECTIONC—C
Fig. 1.6.2. Contour Plot of Cladding Thickness

Variations on Flat-Plate Fuel Element.

47

 
 

ANP PROJECT PROGRESS REPORT

  
 
 

    

1

  

  

  

 

1‘&!0 :

3 S D

W
J

L

 

-.‘v.__:‘ ‘ Ah .1;\;\ 3 " b ‘3
i B e P ST

 
     

 

UNCL ASSIFIED
Y-24583

 

 

Fig. 1.6.3. Hot-Dipped Aluminized Coating on Low Carbon Steel. Etchant: 2% nital.

an intermetallic layer can be made with an
accuracy of 0.0001 in. or greater.

While investigating the problem of measuring
the thickness of aluminum coatings on steel, a
new technique was developed which shows con-
siderable promise of eliminating the inaccuracies
introduced by lift-off or poor coupling between
the probe coil and the part. Previously, thickness
measurements have been made by measuring the
amplitude variations of the impedance of the
probe coil as it scans the part. These amplitude
variations are determined with respect to the
impedance of the probe when placed on a thick
section of one of the component metals of the
system being measured. This method requires
that the surface of the part be flat and that the
probe be in intimate contact with the surface
or have a constant amount of lift-off. Even a
small wvariation in lift-off (0.001 in.) introduces
changes in both the amplitude and the phase
angle of the coil's impedance, which greatly

48

affect the thickness measurement. |t was found
that in measuring the thickness of aluminum on
steel it is possible to eliminate the effects of
lift-off by referring all impedance measurements
to the impedance of the coil in air and then
utilizing the phase change of the coil's impedance
as a measurement of the thickness. With this
technique, lift-off greatly affects the amplitude of
the coil’s impedance but causes almost no change
in its phase angle, as illustrated in the impedance
diagram of Fig. 1.6.4. The phase angle, 8, of
a vector from the air point to the thickness curve
is used as a measure of the aluminum coating
thickness. The introduction of lift-off causes the
vector length to shorten, but has no appreciable
effect on the angle 6. By using this technique
the coating thickness measurements can be made
to be largely independent of the amount of lift-off.

A plot of the phase-angle variation as a function
of aluminum coating thickness on steel is shown
in Fig. 1.6.5. As in all eddy-current thickness

 
UNCLASSIFIED
ORNL—LR—DWG 26994

 

 

 

 

 

o Fe
r—
z 0.3
Ll
=
o
a
=
C
“ 7/
"
> A
5 < 7
o ; P-4
@ : _-
——
o )'/
9‘ 5
Al THICKNESS ON Fe (in. x 1073)
RESISTIVE COMPONENT
Fig. 1.6.4. Normalized Probe Coil Impedance Vari-

ations for Varying Thicknesses of Aluminum on Leow
Carbon Steel at 20 kc Referenced to Impedance of
Probe Coil in Air.

measurements, the most sensitive measurements
are those on the thinnest sections. At the present
time the phase-angle measurement technique can
be used to advantage only if the base material
is magnetic. A study is being made to determine
whether the technique can be successfully applied
to systems having nonmagnetic bases. This
method is particularly well suited to automation
because of its independence of lift-off variations.

LAMB ULTRASONIC-WAVE STUDIES
R. W. McClung

The detection of laminations in thin (]4 in. or
less) metal sections and the nondestructive
evaluation of bond quality in clad structures are

among the more difficult inspection problems.
Radiographic, eddy-current, and dye penetrant
methods are not applicable to these problems

because of unfavorable orientation and non-
accessibility of laminar discontinuities. Con-
ventional pulse-echo ultrasonic techniques do
not have the requiredresolving abilities. Resonance
ultrasonic methods can, in some instances,
provide a partial solution to the problem; however,

PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
ORNL—LR—DWG 26997

 

 

 

 

 

 

 

 

 

 

 

 

140 '\/fi}
.—-—""——-—-—_—-—.
% 100 [ - —pf—
=
}—
w
T 80 _— f— — — -+ —
wn
L
_
g |
Z 60 | — —;— -
U_I !
@ |
<{ |
& 40 — - - ’—— -
20 _ o _ — J,_ ]
|
o 1
O 5 10 15 20 ——=o

THICKNESS OF ALUMINUM ON LOW CARBON
STEEL (in. x 107 3)

Fig. 1.6.5. Eddy Current Thickness Gage Calibration
Curve, 20 ke,

the sensitivity of the resonance method is limited,
and, being a contact technique, it is subject
to the inherent difficulties of contact testing:
relatively slow inspection speed and problems
of contact pressure and application of couplings.

A possible solution to these problems involves
an ultrasonic phenomenon known as Lamb waves,
which are basically a characteristic elastic
vibration established in a thin metal plate.
The vibration has a frequency in the ultrasonic
and the frequency, plate thickness, and
the mode of the vibration are interrelated. This
phenomenon was postulated by Horace Lamb in
1916 on the basis of a mathematical study.

A program of study has been initiated to gain
fomiliarity with this generally unknown technique.
The Lamb waves are being generated in thin
metal sections by directing a beam of ultrasonic
energy at particular incident angles to the sheet
At the present time, commercial ultra-
sonic instrumentation is being used to generate

range,

surface.

and receive these vibrations. As more familiarity
with the system is gained, it is planned to build
specific which will optimize
the test requirements. Early results in this study
have been quite promising.

instrumentation

49

 
ANP PROJECT PROGRESS REPORT

ULTRASONIC ATTENUATION AND VELOCITY
MEASUREMENTS

J. W. Allen R. W. McClung

The evaluation of any ultrasonic inspection
result is difficult because of the attenuation of
the ultrasonic energy in the part being inspected.
This attenuation is almost entirely a function
of the test metal structure and the frequency of
the ultrasonic waves. The problem of ultrasonic
attenuation is particularly troublesome in the
inspection of weldments. In many instances, the
attenuation is so severe that gross weld defects
are not detectable. In order to overcome these
problems of ultrasonic attenuation, much more
information and knowledge must be obtained about
the phenomenon. An ultrasonic attenuation com-
parator has been procured for studying the atten-
vation problem in more detail, and familiarization
studies with this new equipment have begun.

It is planned to study the changes in ultrasenic
attenuation as a function of changes in grain
size, size and dispersions of included matter,
and other changes in metal structures that can
be produced by heat treatment and working.
Another phase of the attenuation studies will
be an attempt to provide valid ultrasonic in-
spection of welded structures.

An electronic counter type of time intervalometer
has been obtained which is capable of measuring
the time interval between successive ultrasenic
pulses to within +0.1 psec. This instrument is
presently being used with an existing pulse-echo
ultrasonic instrument for a very accurate meas-
urement of ultrasonic velocity. |t is planned to
evaluate the changes in ultrasonic velocity as
a function of the parameters noted above for
the ultrasonic attenuation measurements. In
addition, these velocity measurements will be
used in basic studies of the elastic properties of
metals.

 

 
Part 2
CHEMISTRY AND RADIATION DAMAGE

 

 
 

 
2.1. MATERIALS CHEMISTRY

W. R. Grimes
Chemistry Division

PREPARATION OF CHARGE MATERIAL FOR
REDUCTION TO YTTRIUM

G. J. Nessle

Methods have been developed for the preparation
of the LiF-Mng-YF3 mixture used in the
production of yttrium metal and for the delivery
of the mixture into the reduction vessel. The
production process consists in preparation, by
a method similar to that used successfully in
the past for the preparation of fluoride fuel
mixtures, of a suitable mixture of LiF, MgF,,
and YF, that is virtually free of oxygen, reduction
of this mixture with metallic lithium to produce
a low-melting yttrium-magnesium alloy, removal
of the magnesium and the lithium by vacuum
distillation, and consolidation of the resulting
yttrium sponge to a massive ingot. The reduction,
distillation, and consolidation steps are carried
out by the Metallurgy Division in the manner
described in Chap. 1.1 of this report. Details
of the methods used in the production of the
fluoride-mixture charge material by the Chemistry
Division are presented below. Since conversion
of Y,0, and suitable compounds of magnesium
to fluorides can be accomplished on a large
scale in available equipment, a pilot plant
capable of supplying a quantity of fluoride mixture
sufficient for the production of 200 Ib of yttrium
metal per month has been designed and is being
constructed.

Conversion of Y203 to YF3

J. Truitt

Complete conversion of an oxide to a fluoride
by treatment of the solid with anhydrous HF
is usually difficult because the product
fluoride tends to coat the oxide and greatly
diminish the rate of conversion. Since, however,
HF treatment of the LiF-MgF,-YF, liquid is
the final oxygen-removal stage in the yttrium
production process, complete conversion of Y 0,
to YF, in the first step is not essential. There-
fore conditions for the treatment of Y, O, by
HF have been examined in small-scale equipment
similar to the 250-lb-scale equipment which was

vapor

used successfully for conversion of ZrCl, to
ZrF .

The treatment temperatures used have varied
from 1100 to 1400°F ond HF flow rates have
varied between 84 and 132 liters/hr for a 3-lb
charge of Y203. The processing time was usually
12 hr, but one test was terminated after 6 hr.

Evaluation of the product by petrographic
examination, x-ray diffraction, chemical analysis
for fluorine and yttrium, and by measurement of
the oxygen evolved upon treatment with BrF
indicated that the products of all runs were
satisfactory. The conversion, as calculated from
the fluorine-to-yttrium ratio, was above 97% in
all cases and was, in general, above 99%. The
oxygen content was 1.5% in the worst case and
was usually below 0.5%.

Some further testing will be required to establish
optimum processing times and HF flow rates,
but the results show that a temperature of 1100°F
and the lowest flow rate of HF used thus far
are adequate. The scale-up of the process to
that to be used in the very similar 250-lb-capacity
equipment should present no great difficulty.

Preparation of MgF,
J. Truitt

Inquiries are still being made, but thus far
no MgF, which passed all specifications has
yet been offered by commercial sources at a
reasonable price.  Accordingly, conversion of
MgO and Mg(OH), to MgF, by hydrofluorination
of the solid materials has been investigated.
The effect of temperature and HF flow rate have
been examined over the same range as in the
Y,0; hydrofluorination described above. The
hydroxide was found to be more readily converted
than the oxide and to yield an appreciably purer
product. The hydroxide has the disadvantage,
however, that its bulk density is lower than that
of the oxide and therefore less material can be
produced per run in the equipment and higher
losses of powder to the gas exit train occur.
Modifications have been made in the gas exit
line to alleviate the latter difficulty.

53

 

 
ANP PROJECT PROGRESS REPORT

A temperature of 1100°F was found to be
adequate for conversion of either magnesium
compound. Petrographic and x-ray examination re-
vealed traces of MgO in three of four preparations
made from the oxide but showed no such material
in preparations made from Mg(OH),. The con-
version, based on the fluorine-to-magnesium ratio
obtained from chemical analyses, was above 95%
in every case and was, in general, better than
98%. |t appears that sufficiently good conversion
to provide feed for the liquid-phase purification
step can be assured and that scale-up to 250-1b
equipment will not prove to be difficult,

Preparation of the LiF-MgF ,-YF, Charge
Mixture

J. Truitt

In the present process for producing the yttrium-
magnesium alloy used in the preparation of yttrium
metal, a fluoride mixture containing 25 wt % LiF,
21.1 wt % MgF,, and 53.9 wt % YF, is charged
to the reduction vessel. |t is essential that the
fluoride mixture be very pure, especially with
respect to oxide concentration, since the oxygen
content of the metal depends directly on the oxide
concentration of the fluoride mixture. In order
for the metal to have the low oxygen content
specified, the fluoride mixture must not contain
more than 500 ppm oxygen.

In the preparation of the charge mixture, the
individual fluoride salts LiF, MgF,, and YF,
are blended in the proper proportions in a 5-lb-
capacity nickel reaction vessel. The vessel is
purged with helium to remove entrapped air and
moisture and then with anhydrous HF at room
temperature. The mixture is then heated to
1500°F wunder an atmosphere of HF. At this
point, the melt (mp, 1210°F) is subjected to
successive treatments of hydrogen and anhydrous
HF to remove the last traces of oxides and
reducible impurities. At the completion of the
treatment the batch is tronsferred through a
sintered nickel filter and sampler into a nickel
can, - which is so designed that the
fluoride mixture can be transferred directly into
the reduction vessel when needed. The melt
is allowed to cool and the receiver can is stored
on a helium header system under a positive
pressure of helium.

J. E. Eorgan

receiver

A sample of each preparation is submitted for
chemical and petrographic analysis in order to

54

determine whether the purity and the composition
meet the specifications. At the present time,
no satisfactory method is available for determining
the oxygen. content of the purified fluoride melt,
but the analytical method used for determining
oxygen in the YF, is being modified and adapted
for use on these melts (see Chap. 2.2, this report).
|f precise analyses indicate the need, steps can
be taken to change the processing procedure to
lower the oxygen content of the melts.

Thus far, ten batches of the fluoride mixture
have been made with YF, and MgF, processed
from the oxides. Petrographic and x-ray onalyses
have not detected oxides in these batches.

The System LiF-YF,
R. E. Thoma

Magnesium fluoride is present in the fluoride
mixture charged to the reduction equipment only
because the magnesium released upon reduction
drastically lowers the freezing point of metallic
yttrium and permits separation of molten metal
from slag at a readily attainable temperature
(that is, below 1000°C). If the reduction could
be accomplished at a temperature above 1600°C,
the melting point of yttrium metal, the magnesium
would be unnecessary, and subsequent vacuum
distillation and consolidation steps could be
avoided. In this instance, the binary system
LiF-YF; could be used as the charge material
if satisfactory compositions could be melted at
temperatures sufficiently low to permit purification
by the standard method (described above) in
equipment of copper or nickel.

A preliminary investigation of the binary system
LiF-YF, has indicated that liquidus temperatures
in the composition region 15 to 50 mole % YF,
are much lower than those reported in the Russian
literature.!  Results of recent thermal-gradient
quenching experiments are consistent with thermal-
analysis data, to within 5°C, and show the
presence of a single eutectic, containing 23
mole % YF, which melts at 699 +4°C. A previous
visual observation in this laboratory had indicated
a solidus temperature of 690°C in this composition
region.? A single LiF-YF3 compound has been
observed. Optically, it is uniaxial positive with

 

]E. P. Dergunov, Doklady Akad. Nauk S.5.5.R. 60,
1185 (1948).

2R. 1. Sheil, ANP Quar. Prog. Rep. Sept. 30, 1957,
ORNL-2387, p 114.

 

 
an ordinary index of refraction of 1.454 and an
extraordinary index of refraction of 1.472. Results
of quenching experiments show that this compound
contains at least 40 mole % YF_. Although the
data are as yet insufficient for the preparation
of a phase diagram of the LiF-YF3 system, it
is obvious that satisfactory compositions are
available that will
standard equipment.

permit processing in the

EXTRACTION OF LITHIUM METAL IMPURITIES
WITH MOLTEN SALTS

G. M. Watson J. H. Shaffer

Experiments in the selective precipitation of
oxides from molten fluoride mixtures have sug-
gested that lithium oxide is relatively soluble
in such molten salts, It should, accordingly,
be possible to remove lithium oxide from
molten lithium by extraction with a molten
salt, and it may be possible at the same time
to remove lithium nitride and lithium carbide.
Since as a consequence of such a separation
the lithium metal will be saturated with the
extracting salt at the temperature of separation
of salt and metal, it will be advantageous to
use a low-melting salt mixture for the extraction
in order to minimize the salt content of the
lithium, Metallic lithium is such an active
reductant, however, that, if contamination of the
lithium by extraneous metal is to be avoided,
only salts of lithium may be used. These con-
siderations led to the choice of the eutectic
mixture LiF-LiBr (12-88 mole %; mp, 453°C)
for preliminary tests of the extraction method.

Treatment of molten lithium at 600°C with the
LiF-LiBr mixture should extract most of the
lithium oxide from the metal, but the salt mixture
will be saturated with metallic lithium. Therefore,
in preliminary experiments, metallic bismuth has
been used in an additional step to extract the
dissolved lithium from the salt. If the lithium
is completely removed, titration of an aqueous
solution of the salt should provide a convenient
measure of the lithium oxide removed from the
molten lithium.

The LiF-LiBr eutectic was prepared by adding
LiF that had been purified by hydrofluorination
and hydrogenation to dehydrated LiBr. The
bismuth extractant was prepared by reducing
commercial-grade bismuth with hydrogen at 600°C
in a graphite-lined nickel reactor and transferring

PERIOD ENDING MARCH 31, 1958

it through a stainless steel filter disk of 0.0008-
in. porosity in a stainless steel line to a storage
container; the stored bismuth was then treated
with additional hydrogen at 400°C wuntil used.

The lithium metal purification apparatus consists
of an extraction chamber and auxiliary storage
containers. The entire apparatus, including lines,
is constructed of stainless steel; it is filled
with a hydrogen 4tmosphere and heated to 700°C
prior to use. An argon atmosphere is maintained
at all times aofter the hydrogen-firing step.

The purification procedure consists in trans-
ferring, to the extraction chamber, a portion of
LiF-LiBr salt mixture measured by differences
in liquid levels in the salt storage container.
The lithium to be purified is then transferred to
the extraction chamber from o weighed storage
container.  All transfers are effected by pres-
surizing with argon, and changes in transfer-line
connections are made inside a small, inert-
atmosphere dry box. The two-phase mixture is
then agitated for 1 hr at 600°C by energizing a
magnetic coil. (The optimum extraction time has
not yet been determined.) Following the extraction
a dip leg is lowered to the salt-metal interface
and the lithium is transferred to a clean storage
container.

An extraction of dissolved and entrained lithium
from the salt is made at 600°C by repeating the
extraction procedure with liquid bismuth as the
extractant. Following this, a sample of the salt,
taken by means of a filter tube, is dissolved in
water and its total alkalinity is determined by
titration with standard acid. The alkalinity of
the pure salt mixture having been previously
determined, a calculation of the quantity of
lithium impurities removed from the metallic
lithium is then possible. One experimental
determination has been completed, and the results
are presented below:

Weight of salt 1387 g
Weight of lithium metal 675 g
Salt-lithium extraction time 1 hr
Salt-bismuth extraction time 34 hr
Lithium impurities extracted 324 ppm

{caleculated as 02)

The effectiveness of these methods for purifi-
cation and analysis of the purified product cannot,

55

 
ANP PROJECT PROGRESS REPORT

of course, be evaluated on the basis of this one
experiment. It is planned to test the methods by
repeating the purification procedure with the same
batch of lithium. If the method is effective, the
amount of impurities removed as the number of
extractions increases should rapidly diminish.

When purification appears to have been attained,
known amounts of an oxide, probably CuO, will
be added to the purified lithium metal, and the
process will be repeated to determine its reliability.
The apparatus will then be modified to permit
sampling and onalysis of the metallic lithium.

 

 
 

PERIOD ENDING MARCH 31, 1958

2.2, ANALYTICAL CHEMISTRY

J. C. White
Analytical Chemistry Division

DETERMINATION OF OXYGEN IN METALLIC
LITHIUM

A. S. Meyer, r. R. E. Feathers

The precision of a method which has been
proposed for the determination of lithium oxide
in metallic lithium' was evaluated by analyzing
samples of a dispersion of metallic lithium in
paraffin. Lithium oxide was assumed to be
uniformly distributed with respect to the metal
in the solidified hydrocarbon, and therefore a
source of standard samples of unknown, yet
vniform, oxide concentration was available.

The dispersions were prepared with a Waring
Blendor fitted with a stainless steel mixing cup
in which a helium atmosphere was maintained.
In this cup, 95 g of lithium was combined with
230 g of paraffin. The lithium used had been
pretreated, in an effort to reduce the concentration
of dense impurities, by centrifugation of the
molten metal ot a temperature slightly above its
melting point. The metal was dispersed in the
paraffin by heating the contents of the vessel to
230°C ond agitating it vigorously for 30 min. The
dispersion was then cooled rapidly to avoid
segregation of the oxide.

Samples of the solidified material that contained
approximately 1 g of the dispersed metal were
analyzed.  The analytical procedure described
previously2 was followed in detail, except that
the samples were wrapped with perforated silver
foil to moderate the rate of dissolution of the
metal. The results of the analyses are presented

below:

 

TA. S. Meyer, Jr., et al., ANP Quar. Prog. Rep.
Sept. 10, 1956, ORNL-2157, p 127.

2A. S. Meyer, Jr,, and R. E. Feathers, ANP Quar.
Prog. Rep. Sept. 30, 1957, ORNL-2387, p 149,

3A. S. Meyer, Jr., and G. Goldberg, ANP Quar. Prog.
Rep. Sept. 30, 1957, ORNL-2387, p 150.

4, Sheft, A. F. Martin, and J. J. Katz, J. Am. Chem.
Soc. 78, 1557 (1956).

 

Lithium (g) Oxygen (ppm)
0.76 240
0.90 243
0.97 222
1.01 224
1.02 220
Average 230
Standard deviation 11
Coefficient of variation 5%

Since the absolute standard deviation of the
results corresponds to a titration error of only
about one drop of 0.01 N standard acid, the
precision is considered to be satisfactory. Further,
the results indicate that the assumption that
lithium oxide is uniformly distributed in the
lithium-paraffin dispersion is valid.

DETERMINATION OF OXYGEN IN FLUORIDE
SALTS

A. S. Meyer, Jr. G. Goldberg

The use of the addition compound, KBrF4, which
is formed by the reaction of equimolar amounts
of KF and BrF,, as the active reagent for the
determination of oxides in fused mixtures of
fluoride salts has been described in previous
reports.3 This compound, KBrF,, functions as
a basic flux, and, as such, reacts readily with
acidic oxides such as yttrium oxide, but it reacts
slowly with basic oxides such as magnesium
oxide. Sheft, Martin, and Katz? recommend the
use of an acidic flux, Ber-SbFé, which is the
addition compound of BrF ., and SbF ., for reactions
with strongly basic oxides. Accordingly, the
apparatus for the determination of oxides in fused
mixtures of fluoride salts is being adapted for
use with BrF2-SbF6. This flux should be ap-
plicable to the determination of oxides in fused
mixtures of lithium fluoride and magnesium fluo-
ride.

57

 
A

ANP PROJECT PROGRESS REPORT

2.3. RADIATION DAMAGE

G. W. Keilholtz
Solid State Division

CREEP AND STRESS RUPTURE tested in air in order to avoid the changes of .
TESTS UNDER IRRADIATION thermocouple calibration observed earlier when the
). Wilson specimens were tested in helium, and air cooling
W E. Brond W. W. Davi of the specimens will be used to achieve better
ise SRMRAUGS S temperature control. The earlier apparatus was
N. H. Hinkle J. C, Zukas

subjected to temperatures on the order of 1700°F
as a result of gamma heating. Since the use of air
rather than helium surrounding the apparatus would
result in even more severe overheating, a “‘cold
finger'’ that is cooled by circulating air is centered
in each specimen (as shown in Fig. 2.3.1).

MTR Experiments

The tube-burst stress-rupture apparatus for tests
under irradiation in the MTR was completed and
final preirradiation tests are under way. This six-
specimen apparatus is similar to the eight-specimen

apparatus operated le-::1r|iita|',‘l except for two im- V). . Wilscr a al,, ANP Qudv. Prog. Reb. Juse 30,
portant modifications. The specimens will be 1957, ORNL-2340, p 268.

 

UNCLASSIFIED
PHOTO 42859

 

 

 

 

 

 

 

 

 

I[||||||||||||||I]]]l||l||||l|| 3
i 2 3
INCHES
T ——
e

 

Figs 2.3.1. In-Pile Stress-Rupture Specimen and Cooling Means. At the top is o complete stress-rupture specimen
and capillary tube for pressurizing to the test stress. The cold finger projects from the right end. In the middle is a
blind-end, cold-finger tube ready for welding into @ specimen. At the bottom is a hairpin tube for supplying cooling
air to the top of the cold finger.

 

 

 
Cooling of the specimen takes place by radiation
from the inside surface of the specimen tube to
the outside of the cold finger tube. At the in-pile
end of the apparatus where the gamma heating is
greatest, there is also some cooling by metal
conduction. The partially assembled MTR appa-
ratus is shown in Fig. 2.3.2. Electrical heating
was used to simulate gamma heating in out-of-
pile tests made to establish the design con-
figuration. The MTR tests will be operated at
1500°F at stresses from 3000 to 6000 psi. The
specimen tube walls are 0.03 in. thick. Out-of-
pile tests of single specimens in air have given
results that agree with those obtained earlier
by the Metallurgy Division.

Since air is to be used as the specimen environ-
ment in the MTR tests, little more has been done
in the study of the effects of helium and con-
taminants from the experimental apparatus on the
change of calibration of Chromel-Alumel and iron-
constantan thermocouples.?  X-ray diffraction,
Curie point, and metallographic studies have been
started.

 

2). C. Wilson et al., ANP Quar. Prog. Rep. Dec. 31,
1957, ORNL-2440, p 207.

 

PERIOD ENDING MARCH 31, 1958

LITR Experiments

An examination was made of an Inconel stress-
corrosion specimen with a 0,02-in.-thick wall
that had operated in the LITR for 625 hr before
rupture., The stress on the specimen was 2000
psi; the test temperature was 1500°F; and the
specimen environment was NaF-ZrF -UF, (62.5
12,525 mole %) on the inside of the capsule and
helium on the outside. The post-irradiation
examination showed a maximum of 3 mils of
attack at the center of the specimen where the
control thermocouples were located., The attack
was in the form of scattered voids at grain bound-
aries. The attack on the helium side of the tube
was approximately the same in terms of depth,
distribution, and density of wvoids. Rupture
occurred in the neighborhood of the liquid-vapor
interface in the upper part of the capsule. It is
possible that the temperature exceeded 1500°F in
this region.

In common with other in-pile stress-rupture
tests, the regions of the test specimen that showed
the typical severe intergranular separations of
third-stage creep were localized in the neighbor-
hood of the fracture. In out-of-pile tests, the

UMNCL ASSIFIED
PHOTO 42860

 

Fi Qe 21-3-2-

in-pile ends of several specimens are visible. The !fi-in. tubes will supply cooling air to the cold fingers inside

the specimens.

Partially Assembled Stress-Rupture Apporatus Designed for Insertion in HB-3 of the MTR. The

 
ANP PROJECT PROGRESS REPORT

occurrence of the intergranular separations is
more widespread throughout the specimen length,

ORR Tests

The apparatus-holding fixture for attaching
experiment cans to the pool face of the ORR has
been completed and installed, The surfaces of
the reactor tank were neither plane nor well-
and therefore careful shimming was
to obtain the required dimensional

Several sets of the experiment latch-

aligned,
necessary
tolerances.
ing mechanism and the handling tool have been
designed ond built, A sample holder has also
been built that will be used initially to hold
thermocouples for measuring the pool face temper-
ature during shakedown operations. Design work
is proceeding on experiment cans and terminal
facilities for experiment cables. An air-filled can
has been designed and built to fit against the
pool face of the reactor for use in monitoring the
flux pattern in the pool facility. The can will
also be used by the Operations Division to de-
termine the effect of voids on reactivity.

Mockups of conceptual models of in-pile creep
apparatus for irradiation at the ORR have been
tested in order to determine the best specimen
geometry for large numbers of specimens in the
presence of gamma heating that is expected to be
of the order of 10 w/g. The objective is to
accommodate about 18 specimens in a 15in.
length of one of the lattice positions, but it may
be difficult to hold the temperature down to the
desired 1500°F. Instrumentation, safety circuit,
and equipment layouts are being planned.

RADIATION EFFECTS IN
ELECTRONIC COMPONENTS

J. C. Pigg C. C. Robinson
0. E. Schow

The current-voltage characteristics of p-n
junctions have been studied extensively. The
two principal pictures currently used to explain
barrier behavior are the junction theory of Shockley?
and the carrier generation and recombination
1‘heory5 of Pell,* which was later extended by Sah
et al,

 

3W. Shockley, Bell System Tech. |. 28, 435-89
(July 1949).

4E. M. Pell, J. Appl. Phy. 26, 658 (1955).

5C. T. Sah, R. N. Noyce, and W. Shockley, Proc.
Inst. Radio Engrs. 45, 1228 (1957).

60

The junction theory of Shockley predicts a
saturation current in the reverse direction and a
forward current proportional to e7Y/%T,  The
carrier generation theory does not postulate
saturation in the reverse direction and predicts an
e?V/"kT  Jependence in the forward direction.
The former picture can be used to explain the
current-voltage characteristics at room temperature
of semiconductors having small forbidden gaps -
for example, germanium —and the latter to explain

these characteristics in material having large

Pell has

shown that even in germanium the charge-generation

forbidden gaps — for example, silicon.

process predominates at low temperatures.,

The reverse current of a junction may be ex-
pressed as the sum of two terms:

IR = Id + I’_g ’

where

[, = the junction current,

L= the generation current,
The reverse junction current is given by Shockley
as

_ e-qV/kT)

with

= 1/2
1/2
IflS = Aqn (Dfl/Tfl) /

where the diffusion constant, D, and lifetime,
7, are those corresponding to the location in the
crystal where the carrier considered exists as a
minority carrier and where

V = voltage,
g = electronic charge,
n,p = density of electrons and holes, re-
spectively,
Dn,Dp = diffusion constant of electrons and
holes, respectively,

T Ty = lifetime of electrons and holes, re-
spectively, when appearing as a mi-
nority carrier,

k = Boltzmann’s constant,
T = absolute temperature,
A = area of the barrier.

 
The reverse current component from the charge
recombination-generation mechanism as given by

Pell is

I =qGWA ,

where G is the charge regeneration rate, [ is the
fraction of the space-charge region involved in
charge generation, and W is the thickness is the
barrier,

These factors have been given in terms of the
electronic properties of the material as follows:

b, (1 - e—-qV/kT)

G =

 

'rn(n + n]) + ‘rp(p + p]) '

V+e ~2c¢
g T

V+VAF

where n, and p, are the respective carrier densi-
ties that would occur if the Fermi level were at
the energy level of the recombination center,
€1 is the band gap, €_is the depth of the recom-
bination center from the nearest allowed band,
VA is the barrier height.

The barrier thickness, W (in cm), can be de-
termined in terms of the barrier capacitance, C
(in ppf), which for germanium is

Ww=142A/C ,

and n,, p, by definition, are given by

(¢. — € Y/kT
n,=N_e T &

(eg - Er)/kT .
p] =Nve

where N_ is the density of states in the conduction
band and N, is the density of states in the valence

band.

A combination of these factors gives the reverse
current as

PERIOD ENDING MARCH 31, 1958

It can readily be seen that the I, term correspond-
ing to the recombination-generation mechanism
does not indicate saturation in the reverse di-
rection and that this term is also more strongly
dependent upon carrier lifetime than is the dif-
fusion component.

Ifit is assumed that the recombination-generation
level is appreciably closer to one band than to
the other, the expression can be simplified, If,
for example, the level in question is near the

 

conduction band, then Ty > Tb e The ex-
pression then becomes
].42A2 V+ Eg —2€r
I = pd
T8 C V+Vag
Nv (eg - er)/kT
X —e (1 —e-aV/kT) |
,

The Irg term in I, is exponentially related to

(eg_sr)/kT

temperature by the factor e

The
I, term in I, is exponentially related to temper-

-€ _/kT

ature by the factor ¢ & ', It would be ex-

pected, as Pell found, that the [, term predominates
at high temperatures and the Irg term predominates
at low temperatures in a germanium junction.
Curtis et al.® found a recombination level about
0.2 ev below the conduction band, Such a level
justifies the assumption made above. They found
that the lifetime decreased essentially linearly
with irradiation exposure, and they attributed the
decrease in lifetime to the introduction of recombi-
nation centers by the irradiation.
which the
R would be expected
to increase as a result of the decreased minority
carrier lifetime. Since I, = 1/71/2 and 1, 1/,
L. would increase at a faster rate than I, In ad-
dition, new

For small-dosage irradiations in

Fermi level is not changed, I

recombination-generation centers

 

0. L. Curtis, Jr., et al., |. Appl. Phy. 28, 1161 (1957).

 

1.4242 gn p(V + €p = 2€r)

 

L= C [-rn(n +n)+ 'rp(P+ Pl)] [V + VAF]

(1 - e—-qV/kT)

61

 

 
ANP PROJECT PROGRESS REPORT

that will add a new term lr' to I, will be intro-
duced at 0.2 ev below the conduction band.

If 1, is defined from the limit as V goes to
zero of the slope of the reverse current-voltage
characteristic,

{ I il li 4 (1, )
=1,+— lim — .
R d q V_'Odv rg
1/2 1/2
D, D
= Agp|— +Agn | — +
T T
p n

2 n, eg—2€r
+142 —g—|—| .
C

s\ YAF

Curves showing the effect of temperature on I,
for a IN91 germanium-alloy-junction power rectifier
before irradiation, after an integrated fast neutron
exposure of 10'3, and after an exposure of 1074
are presented in Fig. 2.3,3. Before irradiation,

UNCLASSHFIED
— 4 ORNL—LR—DWG 30433

 

! ~ 0,2ev
|

1077 ‘ | \
A BEFORE IRRADIATION \
o INTEGRATED FAST NEUTRON
FLUX OF 10"
o INTEGRATED FAST NEUTRON

\ FLUX OF 10"

t\ 0.54 ev

10°7 \

ACTIVATION }
ENERGY, O.75ev

 

 

 

 

—6

 

I, (amp)
o

 

 

 

 

 

 

 

 

 

 

 

—8

10

340 360 380 400 420 440 460 480
103/7 (°K)

 

Figs 2.3.3. Curves Showing Effect of Temperature on
Iy for IN91 Germaonium-Alloy-Junction Power Rectifier
Before and After Irradiation in ORNL Graphite Reactor.

62

the slope of the line shows an activation energy of
0.75 ev, which corresponds to € . This indicates
that the Shockley diffusion theory applies to this
sample. After irradiation in the animal tunnel of
the ORNL Graphite Reactor to an exposure of 10'3,
lo had increased by about a factor of 4. Measure-
ments of barrier height showed no appreciable
change, and thus no appreciable change in the
Fermi level was indicated. The curve does
not show a sharp break of the kind reported
by Pell: in fact, the slope seems to increase
at low temperatures. It is believed at present
that this increase in slope is experimental
error resulting from the small currents involved
in measuring [ at low voltages. After an ex-
posure of ]0]4, the curve assumed a slope of
approximately 0.2 ev, in agreement with the ex-
pectation of a predominating I _ as a result of a
recombination-generation level 0.2 ev below the
conduction band.

The absence of saturation in the reverse current-
voltage  characteristics curves of germanium
diodes after the units have been exposed to
irradiation is thus due to a change in rectification
mechanism  resulting from the bombardment-
produced recombination-generation centers. Such
a change in mechanism will also change the
forward characteristic. Hence, changes in the
emitter behavior of transistors subjected to radi-
ation may be expected.

IRRADIATION TESTS OF MODERATOR
MATERIALS FOR USE AT
HIGH TEMPERATURES

W. E. Browning R. P. Shields
J. E. Lee, Jr.

Preparation of apparatus, as previously de-
scribed,” for irradiation of moderator materials at
high temperatures in the ETR has been continued.
The yttrium hydride capsules submitted by GE-
ANPD for this experiment were examined radio-
graphically and were found to have large hidden
radial cracks. Additional yttrium hydride speci-
mens have therefore been requested; they will be
encapsulated at ORNL. The other specimens for
the initial test are ready.

The specimens are right cylinders, and they
will be stacked in the C-33 H10 position in the

 

w. E. Browning and R. P. Shields, ANP Quar. Prog.
Rep. Dec. 31, 1957, ORNL-2440, p 209.

 

 
 

ETR in the order shown in Table 2.3.1, which is
the order from the top down with the center line
of the reactor between numbers 3 and 4. Table
2.3.1 also gives the specimen size, material, and
gamma heating it will receive. Each capsule
will contain three specimens,

Table 2.3.1. Moderator Materials for
Irradiation in the ETR

 

Height of

 

G
Capsule Each 1-in.«dia Moderator amma
) Heating
Number Specimen Material
. (w/g)
(Inv)
1 0.5 Beryllium oxide 18
2 0.75 Bery llium 23
3 0.75 Yttrium hydride 25
4 1.0 Yttrium hydride 25
5 1.0 Beryliium oxide 23
6 1.0 Yttrium hydride 18

 

PERIOD ENDING MARCH 31, 1958

Compatibility tests are being run to determine the
effect of yttrium hydride on Inconel and on type
446 stainless steel when heated in contact, in a
hydrogen atmosphere to a temperature of 900°C,
The duration of the test will be several weeks.

Several mockup experiments have demonstrated
that the heat from the hottest specimen can be
carried across a 10emil annulus filled with a
helium-nitrogen mixture which will be essentially
static. Also it has been demonstrated that thermo-
couples can be attached to the walls of the
capsule in the 10-mil annulus,

Several preirradiation tests have been or are
being made on all specimens. The yttrium hy-
dride specimen will be studied by x-ray diffraction;
it will be heat cycled; and measurements will be
made of its elastic properties and dimensions.
The other specimens will be studied similarly,
except that they will not be examined by x-ray
diffraction. The x-ray diffraction examination of
yttrium hydride wili be made to determine crystal
type and lattice spacing. These studies will be
repeated after irradiation in order to detect
changes, if any.

63

 
Part 3
ENGINEERING

A. P. Fraas A. L. Boch

Reactor Projects Division

 
 
3.1. COMPONENT DEVELOPMENT AND TESTING

H. W. Savage

E. Storto

Reactor Projects Division

LOOP FOR INVESTIGATING CORROSION IN
LITHIUM-COLUMBIUM SYSTEMS

D. R. Ward

A forced-circulation loop has been designed, as
shown in Fig. 3.1.1, for testing corrosion in
lithium-columbium systems. The loop will be con-
structed of columbium, and lithium will be circu-
lated with the use of motor-driven electromagnets.
The loop is designed to minimize the need for
cladding to provide oxidation protection of welded
joints. Clad joints are required only at filling- and
expansion-tank connections in the coolest part
of the loop. The welded columbium tubing joints
in heated and cooled parts of the loop will be
surrounded by a liquid metal (sodium) heat-transfer
fluid or an inert gas. High-quality columbium
tubing is presently becoming available only in
short random lengths, and many joints may there-
fore be required in the heated and cooled portions
of the columbium loop. Prior to construction with
columbium, the performance characteristics of the
loop will be determined in an identical loop con-
structed of stainless steel in which NaK will be
circulated.

PUMP DEVELOPMENT
¥W. F. Boudreau A. G. Grindell

Irradiation Test of Bearings and Seals
D. L. Gray

Operation of the rotary element of an oil-lubricated
centrifugal pump installed in a gamma-irradiation
facility inthe MTR canal, as previously described, !
was resumed after the damaged shaft had been
repaired. Examination of the shaft indicated that
the lower journal bearing had been damaged and
that the damage had probably occurred as a result
of temporary, accidental operation of the shaft at
an excessive speed. By late in the quarter, the
bearing and seal parts had been subjected to a
total accumulated gamma-ray dose of 1.6 x 107 rep.
About 98% of the lubricant (Gulfcrest-34) is ex-

ternal to the radiation field, however, and the

 

]D. L. Gray and W. K. Stair, ANP Quar. Prog. Rep.
Sept. 30, 1957, ORNL-2387, p 32.

gamma-ray dose to the bulk of the oil has been
only 3 x 107 rep. Only the lubricant that remains
between seal faces or has accumulated in the
catch basin has received the full dose.

it has been estimated that by the time the bulk
of the oil has accumulated a dose of 5 x 108 rep,
a 25% increase in viscosity will have occurred.
The increase in the viscosity of the lubricant and
the resultant bearing overheating and increased
seal leakage are expected to provide information
with which to evaluate the useful life of the rotary
system in a radiation field. The data obtained
can then be used in the design of elements and
shielding with which to lengthen the useful life.
The effect of irradiation on the physical properties
and composition of the oil will be determined.

Seal Improvements
D. L. Gray

Two modifications made to improve the primary
face seal of a centrifugal pump for high-temperature
service are being tested. @ The modifications
were made to reduce the rate of accumulation of
grease-like materials on the gas side of the seal
where lubricant that leaks through the seal mixes
with gases laden with constituents from the liquid
being pumped. In one modification, the leakage of
lubricant is reduced by the use of a Fulton-Sylphon
seal. In the other modification, purge gas is
admitted to the shaft clearance annulus between
the seal and the liquid being pumped so that flow
of the gas toward the liquid reduces diffusion of
undesirable vapor to the seal region,

The Fulton-Sylphon seal was installed ina liquid
metal pump and had operated about 3700 hr by the
end of the quarter at pump speeds ranging from
2400 to 3450 rpm. The rubbing speed between
seal faces varied from 39.25 to 56.5 fps. Oil
3 Thus
far no evidence of accumulation of deleterious
deposits on the gas side of the seal has been

leakage averaged less than 1 cm® per day.

noted.
The split-purge arrangement, described pre-
viously,? had operated about 3400 hr when the test

 

2D. L. Gray, ANP Quar. Prog. Rep. Dec. 31, 1957,
ORNL-2440, p 25.

67

 
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED

ORNL-LR-DWG 30434
MOTOR-DRIVEN

\ ELECTROMAGNETS
=

7~ LITHIUM EXPANSION TANK
J

N A
— T
N -

HELIUM l
INLET

 

 

 

 

\THERMOCOUPLE LEADS
i’

i‘g

SODIUM EXPANSION TANK A ‘
(HELIUM ATMOSPHERE)

        
 

  

CLAMSHELL HEATERS

& A HOT-LEG LITHIUM-FILLED TUBING
COOLING WATER 7 )
1

INLET
&

 
  
     

 

COLD-LEG LITHIUM-FILLED TUBING —

COOLING WATER
OUTLET
LITHIUM FILL AND x
DRAIN TANK

   

HELIUM PRESSURIZING ‘!
INLET ‘ ’

 

   

SODIUM FILL AND

DRAIN TANK

Fi gs 3.1-]-
Circulating Lithium.

Diagram of Forced-Circulation Loop for Corrosion Testing of Columbium Tubing Exposed to

 
was suspended to find the cause for increasing
restriction to the flow of the gas. Deposits had
accumulated in the shaft annulus and closed the
gas entry ports. Since there were no deposits
above the gas inlets or in any part of the seal
region, it was evident that the purge gas had
successfully stopped the rise of vapor from the
pumped fluid. The cleanliness of the ocil-leakage
catch basin and the deposits in the shaft annulus
may be seen in Fig. 3.1.2, An elastomeric seal
between rotating parts had hardened and it may
have allowed lubricant to leak into the shaft
annulus independently of the action of the pump
primary face seal.

Cavitation Tests
P. G. Smith W. E. Thomas

Experiments have shown that cavitation-induced
damage of an impeller is primarily a function of
the head-vs-flow characteristics of the pump. The
density of the fluid being pumped affects only the
pressures, and therefore pump cavitation tests
with water and liquid metals will indicate the
operating conditions to be expected in the pumping
of lithium with similar pumps. Two tests are now
under way in which pumps are being operated over
a wide range of conditions in order to evaluate
cavitation-induced damage as a function of time.
One pump has operated under cavitation conditions
for more than 4000 hr without changes in per-
formance characteristics and the other has operated
500 hr. Upon completion of these tests, which are
scheduled for a total of 10,000 hr, the impellers
will be examined for cavitation damage.

VALVES FOR USE AT HIGH TEMPERATURES
. T. Dudley

Satisfactory high-temperature operation of a
general-purpose poppet valve with a bellows-
sealed stem has been demonstrated. Valves with
the plug and the seat of Stellite 6 overlaid on
Inconel have been tested with NaK, and valves
with Kennametal cermets® for both the plug and
the seat have been tested with fused salts. Since

 

3The Kennometol cermet designated KM was used for
the plug and the Kennametal cermet designated K-1628
was used for the seat; see ANP Quar. Prog. Rep.
March 31, 1957, ORNL.-2274, p 49, for compositions of

these cermets.

PERIOD ENDING MARCH 31, 1958

cermets of the type tested are known to be com-
patible with lithium, the valve with the cermet
plug and seat would be expected to be satis-
factory for use with lithium. The available valve
bodies were constructed of Inconel and conse-
quently could not be tested with |ithium,

The valve with the Stellite 6 plug and seat was
tested with NaK at 1000°F under a pressure of
50 psi. Leakage decreased from 0.75 cm3/hr to
0.2 cm3/hr after 1500 hr. The force needed to
open the valve remained less than 50 Ib, and a
1300-1b force was needed to effect closure. Post-
test examination showed slight creep of the seat-
ing surfaces, but the closure had not been affected.

The valve with the K-162B cermet seat (Rockwell
A hardness, 89) and KM plug (Rockwell Ahardness,
91) operated satisfactorily when the test liquid was
free of particulate matter. Particulate solids
suspended in the liquid prevented full seating of
the very hard, carefully lapped surfaces of the
mating cermets.

THERMAL STABILITY TESTS
OF METAL SHELLS

J. C. Amos R. L. Senn

The second thin-shell test model was subjected
to a creep buckling test at 1500°C in an atmosphere
of helium with an external pressure of 52 psi.
The shell buckled as shown in Fig. 3.1.3 in 111.15
hr.  The top half of this shell where buckling
occurred had previously been subjected to 339
thermal cycles in a thermal-cycling test stand.
A detailed description of this thermal-cycling test
was presented previously.?

Two additional thin shells of the same geometry
are being fabricated by welding together over-size
sections and reducing the weldment to the desired
final dimensions by machining.  Metallurgical
observation has shown that in some locations the
weld metal that remains after the machining may
consist of a single grain that extends across the
total shell thickness. In order to investigate
further the durability under thermal cycling of a
vesse| fabricated as a machined weldment, one of
the new shells is currently being assembled in
the thermal-cycling test container.

 

4). C. Amos and L. H. Devlin, ANP Quar. Prog. Rep.
March 31, 1957, ORNL.-2274, p 50; J. C. Amos and
R. L. Semn, ANP Quar. Prog. Rep. Sept. 30, 1957,
ORNL.-2387, p 45.

69

 

 
0L

UNCL ASSIFIED
PHOTO 30754

 

Figs3.1.2. Oil-Leakage Catch Basin ond Outer Wall of Shaft Annulus of Centrifugal Pump Operated with o Split=Purge Arrangement to Prevent the Rise
of Vapor from the Pumped Fluid.

 

130d3Y S53390dd L23rodd dNV
 

PERIOD ENDING MARCH 31, 1958

UNCL ASSIFIED
PHOTO 0914

 

Fig. 3.1.3. Thin Shell Weldment That Buckled After 111.15 hr at 1500°F in en Atmosphere of Helium and an
External Pressure of 52 psi.

71

 
ANP PROJECT PROGRESS REPORT

3.2, HEAT TRANSFER STUDIES

H. W. Hoffman

Reactor Projects Division

STUDIES OF THE EFFECT OF
THERMAL-STRESS CYCLING
ON STRUCTURAL MATERIALS

Pulse-Pump System
J. J. Keyes A. |. Krakoviak

Three additional tests (tests 6, 7, and 8) have
been run with the high-frequency thermal-cycling
pulse-pump loop! in which the fuel mixture
NQF-ZrF4-UF4 (56-39-5 mole %) was the circu-
lating fluid and helium was the pulsing and cover
gas. The test sections were 0.485-in.-I1D, 0.415-
in.-wall Inconel cylinders machined from bar
stock.

The operating conditions and results of tests
6 through 8 are given in Tables 3.2.1 and 3.2.2.
This series of experiments has been primarily
concerned with the effects of exposure time (or
total number of cycles) on thermal cycling, Thus,
runs 5 and 6 and runs 7 and 8 each constitute
time-comparison test-pairs which were performed
under conditions as nearly identical as could be
maintained with long-term operation of the experi-
mental equipment. In all tests the fluid flow was
kept constant at 6 gpm (equivalent to 5 = 3700
Btu/hr-ft2.°F in the test section); the pulsing

 

]J. J. Keyes, A, |. Krakoviak, and J. E. Mott, ANP
Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 54.

frequency was 0.4 cps. The stream-temperature
amplitudes were different in the two test groups.

In run 6, incipient intergranular cracks were
found after 22.8 hr of pulsing with a stream-
temperature amplitude of +242°F. This is illus-
trated in Fig. 3.2,1 which shows surface cracks
to a depth of 2.5 mils. The test section of run 5,
as previously reported,? had cracks as deep as
72 mils after 135 hr of pulsing.

For tests 7 and 8, the stream-temperature
amplitude was +139°F, with run 7 lasting 155 hr
and run 8, 351 hr. The results of these tests
are given by the photomicrographs of Figs.
3.2.2 and 3.2.3. The specimen used in test 7
showed variable subsurface void formation (ranging
from light to heavy in concentration) to a depth
of 3 mils. No cracks were found in the test
section. This is to be compared with run 8
(exposed to 2% times as many thermal cycles)
in which massive cracking occurred. Typical
cracks which formed near the test-section entrance
(region just beyond the conical inlet) are shown
in Fig. 3.2.2. The cracks penetrated the 0.415-in,
wall to a depth of 0.080 in.

A section of l-in. sched-40 pipe ]/2 in. down-
stream from the Y-section (junction of the hot
and cold streams) was removed after a total

 

21bid., Fig. 1.7.2b, p 57.

Table 3.2,1. Operating Conditions of HighFrequency Thermal-Stress<Cycling Tests of inconel Pipe

 

Test Variables

Test Number

 

 

6 7 8
Average fluid temperature, °F 1398 1410 1409
Maximum temperature difference between 484 278 278

hot and cold streams, °F

Frequency of temperature oscillations, cps 0.4 0.4 0.4
Total pulsing time, hr 22.8 156.0 352.0
Total cycles 32,800 224,000 508,000
Total fluid flow rate, gpm 6.2 6.3 6.3

 

72

 
Table 3.2.2, Results of High-Frequency Thermal-Cycling Tests of Incone! Pipes

 

€L

 

Wall Measured Temperature  Estimated Temperature .
Test Sample® Thickness Amplitude on Outside Amplitude on Inside Resuits of Metallographl‘cq[ Examination of

No. (in.) Pipe Wall (°F) Pipe Wall (OF)b Test Piece and Adjacent Piping

6 a 0.133 *17.5 %55 Moderate to heavy void formation to a depth of
2 mils above weld

6 b 0.415 Not measured 1136 Cracks to a depth of 3 mils occurred at point where
inside diameter of pipe was reduced to 0,485 in.

6 ¢and d 0.415 Not measured 1136 Cracks occurred throughout test section with severity
decreasing markedly below middle®

6 e 0.415 Not measured 1136 Cracks observed (less severe than at entrance) in
0.485-in.-ID section at point where inside
diameter increased to 1.049 in.

6 f 0.133 +28.0 155 Light void formation

7 Q 0.133 11,0 +32 No evidence of cracks

7 b,c,d 0.415 0 +78 No evidence of cracks; light to heavy voids to a depth of 3 mils

7 e 0.133 £16.0 32 No evidence of cracks in pipe above or below weld; however,
cracks to a depth of 25 mils developed in weld-base metal
interface

8 b 0.133 112.7 132 tntermittent light veid formation to a depth of 3 mils

8 c 0.415 Not measured +78 Very heavy cracking to a depth of 85 mils ‘4 in, below point
where test section is reduced to 0.485 in. ID

8 d 0.415 Not measured 178 Transverse section from center to test piece showed less heavy
cracking than sample ¢; cracks to a depth of 42 mils

8 e 0.415 Not measured 178 Cracks to a depth of 16 mils in small-diameter section that
finally degenerate to moderate intergranular void formation
in diverging regien

8 f 0.133 120.4 132 Light void formation to a depth of 4 mils; no evidence of

cracks in welded zone

 

%For sample location, see Fig. 1.7.1, p 56, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440,

bAssuming 100% efficiency, using Jakob’s equation for calculation of N, and neglecting entrance effects,

c . . . . . .
Some grain-boundary relief was seen in conjunction with some of the well-developed cracks,

before failure.

This was probably due to deformation occurring just

8561 ‘L HOYVW ONIGN3 d0oly3d

 
ANP PROJECT PROGRESS REPORT

 
    

 

UNCL ASSIFIED
T-14514

W

|—u_|--1
I
0
=
0.0Z

CI.D'!‘
<

— O -
21

 

 

Figs 3.2.1. Effect of High-Frequency Thermal Cycling on Inconel in an NuF-ZrF‘i_UF‘ (56=39-5 mole %)
Environment — Test 6 Etchant: Marble’s reagents 100X. (Secret with caption)

exposure of approximately 1200 hr. Of this
total, 408 hr were at a stream-temperature ampli-
tude of +240°F, 606 hr at +135°F, and 183 hr
isothermal ot 1300°F. The section was subjected
to 4.3 x 108 thermal cycles over this period.
Photographs of the sample showed no evidence
of intergranular cracks. Heavy subsurface void
formation to a depth of 3.5 mils was observed.
The lack of cracking may be attributed to (1) a
low heat-transfer rate and hence relatively small
surface-temperature fluctuation ( < 0.25) in this
section, (2) a low cyclic frequency (0.4 cps), for
which the pipe wall is effectively thin, and (3)
stresses that were low over most of the accumu-
lated time.

Thermal Entrance Studies with Water
J. J. Keyes

Measurements of surface attenuation reported by
Mott3*4 for conditions far downstream were ex-

74

tended to measurements of regions close to the
entrance of a thick-walled Inconel cylinder. These
measurements are of interest in connection with
the high-temperature thermal-cycling tests, since a
substantial fraction of the test section is in o
thermal entrance region where increased heat
transfer and thus increased surface temperature

amplitude prevails.

The entrance geometry under investigation
consists of a sudden contraction from a 24-in.
length of l-in. sched-40 pipe (ID = 1.049 in.)
to the shorp-edge entrance of the 0.460-in.-ID
test section, with a consequent area reduction

 

3y, E. Mott, An Investigation of Thermal Transients
11:!9; ifinffd—Ffm'fl’ Interface, ORNL CF-57-10-122 (Oct. 25,
7).

4.!. E. Mott and A. G. Smith, Jr., ANP Quar. Prog.
Rep. Dec. 31, 1956, ORNL-2221, p 54.
- UNCL ASSIFIED
] T-143%6
f . “‘:\. 5 il .' T

.;‘.':.{“fi A

 

Figs 3.2.2. Effect of High-Frequency Thermal Cycling on Inconel in an HQF—IIF‘—UF‘ (56 =39~5 mole %) Environment — Test 8. Etchant: Marble's
e reagents (@) 75X; (&) 100X« Reduced 4%. (Secret with caption)

8561 ‘LE HOYVW 9NION3 goly3d

 

 
ANP PROJECT PROGRESS REPORT

 
 

UNCL ASSIFIED
T=-14391

  

 

 

 

 

 

 

 

|
'l ik
i, / .:-fi s o . e 1?‘;"
{ 3 ¥ od
\ Jl . .
"“'}- N ’ . /_

Fige 3.2.3. Effect of High-Frequency Thermal
Cycling on Inconel in on NnF—ZrF“_UF‘ (56-39-5
mole %) Environment — Test 7. Etchont: meodified
aqua regias 250X. (Secret with caption)

of 5.2 (see Fig. 3.2.4). Greater reduction was
not deemed practical, since the length of pipe
upstream required to establish flow ahead of
the contraction would have resulted in severe
attenuation of the thermal oscillations at the
test section.

Surface temperatures were measured by means
of the Inconel-nickel ‘‘gun-barrel’’ thermocouple
previously described®*4 and stream temperatures
by a 0.005-in. Inconel-nickel wire couple immersed
with the junction approximately at midstream.
Reynolds moduli were varied from 50,000 to
133,000 and x/d from 65 to 4.3 (x is the distance
of the wall thermocouple from the entrance of
the test section of diameter 4d).

Some of the results are plotted in Figs. 3.2.5
and 3.2.6, where the ordinate is the experimental
value of the local attenuation ratio

surface temperature amplitude at x

 

Ne = o :
*  fluid stream temperature amplitude at x

Curves are drawn for frequencies of 1, 2, and
3 cps. The dashed lines in Fig. 3.2.5 represent
calculations based on the Jakob equation,® and

76

UNCLASSIFIED
ORNL-LR-DWG 30435

t-in. SCHEDULE-40
PIPE — TO TEST SECTION

X

SINUSQIDAL FLUID TEPERATURE
INPUT FROM PULSE PUMP

WATER FLOW RATE, 16 gpm
AVERAGE TEMPERATURE, 80°F

ENTRANCE _~WALL "GUNBARREL"
/" THERMOCOUPLE

 

    

 

~S3TREAM THERMOCOUPLE

A -in-WALL INCONEL
TEST SECTION

Fige 3.2.4. Test Section Entronce Geometry.

UNCLASSIFIED
ORNL—LR—DWG 3043

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T T 1 B T I ! ']
08— | ——EXPERIMENTAL REsuLTs | |
----- ESTIMATED FROM ENTRANCE DATA
07 = ' FOR AR, Mg, = 50,000, INFINITE |
| | CONTRACTION (REF 3) i
| L ¥l Bl
! BEENE - EERRE
F:tofi —.4—_—.—.—;— ---?‘TT-!‘.""L?-"'—‘?:!—'-E-E'-'-"I!'-'='—“':-‘-'-‘H-‘-'-":‘—.— :
' el ! ! e
g S 2 N, Y. 00 S U Y
o4l ——— ‘3‘--4‘:;;——4—— 1T 1t
N L T e e siam |
] f=3cps | | |
03— — Ij ' il 11
i AN ] P
02 | | i _| ! | | |
{ 2 5 10 20 50 100

Fige 3.2.5. Variation of Local Surface Attenuation
Ratio, 77, with Distance from Sharp-Edge Entrance in a
Circular Tube of Diameter 0.46 in. at NRe = 47,000
(b, = 2800 Btu/hrft?°F).

UNCLASSIFIED
ORNL-LR-DWG 30437

 

{ 2 5 10 20 50 100
xfd
Fige 3.2.6. Variation of Locaol Surface Attenuation,

M.+ with Distance from Sharp-Edge Entrance in @
Circular Tube of Diameter 0.46 in. at Npo = 133,000

(b = 4870 Btu/hrf+2°F).

 
 

were obtained by replacing the local heat-transfer
coefficient, h,, by the quantity

b

x
Hom,0\ 7~

® / oair

where (b /b ) is obtained from the measure-
ments of Boelfer et al.% for an infinite contraction
at Np, = 30,000, and H, H,0 is determined by

the Dn"rus-Boelfer correlaflon,
_ 0.8 »;0.35
”w,uzo d/k = 0.023 N2 Np-3°

It is apparent that, for the conditions of this
study, entrance effects are not significant above
an x/d of about 6.

The variation of n at x/d = 4.3 for N Neo of
133,000 and 50,000 is shown in Fig. 3.2.7 and
compared with curves (dashed) calculated from
the Jakob equation and entrance data for air,
as described in the preceding paragraph. Note
that the calculations underestimate the effect
of Ngp, and consequently of b on 7, It is in-
tended to continue these measurements down to

 

S, M. K. Boelter, G, Young, oand H. W. lversen, An
Investigatien of Aircraft Heaters XXVI - Distribution
of Heat-Transfer Rate in the Entrance Section of a
Circular Tube, NACA TN-1451 (July 1948),

UNCLASSIFIED
ORNL-LR-DWG 30438

 

 

 

 

 

 

 

 

 

9. ’
0.7 \\\ \ =133 000
\l\ ® 2 o,
\\\ \hm— 4870 Btu/hr- ft<-°F
0.6 S~
‘ [P
--..__.__--‘- \
- n\\ ]
> 05 N - -
Y Y\.\\ e = 50,000
® % ~ = . f+2.9
2 \ x\\/}: 2200 Btu/hr - ft<-°F
e 0.4 e
™~ ~—
\;
0.3 ™~ ‘ i

 

 

 

 

 

 

 

— EXPERIMENTAL RESULTS

0.2 — —]
——— ESTIMATED FROM ENTRANCE DATA FOR AIR,

Npe= 50,000, INFINITE CONTRACTICN (REF 3)
| l i | | I

 

 

 

o1
q 2 3 4
FREQUENCY (cps)
Fig. 3.2.7. Variation of Surface Attenuation, 7, with

Frequency at x/d = 4.3,

PERIOD ENDING MARCH 31, 1958

at least x/d = 1 and to study other entrance
configurations of interest.

THERMAL STRUCTURE OF FLUIDS FLOWING
IN VOLUME-HEATED SYSTEMS

Diverging Annular Channel with Screens
N. D. Greene

The stydy of the effect of screens in a diverging
annular channel on the thermal structure of a fluid
with internal heat generation has been completed.
Three screens with increased porosity in the
region of the outer wall were installed in the
positions shown in Fig. 3.2.8, The topmost
screen was of uniform porosity. A 3.8-in.-thick
drilled Boltaron sheet located in the channel
entrance converted the swirl entrance flow to

A
ORNL-LR-DWG 30439

\ /FLUID IN

 

 

 

 

 

 

 

       
  
  
 

 

 

 

 

[ eorwaros
<!
<
< 8
8 zm SCREEN NO. 1
S -
o
o®
SV s ASCREEN NO. 2
: ‘
©
WO
e SCREEN NO. 3
1 ey s SCREEN NO. 4

INNER WALL

QUTER WALL

Fig. 3.2.8. Location of Screens in Diverging Annular
Channel.

77

 
ANP PROJECT PROGRESS REPORT

axial flow.
Table 3.2.3.

Both steady-state and fluctuating temperatures
were measured at the inner and outer channel
walls. The transient data are compared in Fig.
3.2.9 with data obtained in previous measure-

The screen porosities are listed in

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
 

 

 

 

 

 

 

 

 

ments®+? on the vaned entrance system and on
L ]
ORNL-LR-DWG 30440
oo T

Z UNIFORM SCREEN,

= 0.5 -OUTER WALL

3 oyl

E -~ T~ - VANED - FLOW,

g 04 P }/\ INNER WALL [ |
L_Ll , \“

w 0.3 La” .

: e TS
202 - GuTER VARIABLE SCREEN, QUTER WALL —|
G e O] VARIABLE SCREEN, INNER WALL

Y UNIFORM SCREEN, INNER WALL __|
o - | e o i« ' - . . - —
il T

o b= p s s LT T T ST o L T T
6 7 8 9 {0 1 12 {3 44 15
AXIAL DISTANCE FROM INLET (iny

Figs 3.2.9. Transient Temperature Fluctuations for

Vaned-Flow and Uniform and Variable Porosity Screened
Channels with Two-Pump Operation. Power generation

uniform at 120 kw; NRe = 81,500,

the uniform-porosity screen system. The amplitude
of the surface-temperature fluctuations was re-
duced to less than 10% of the axial rise at both
the inner and outer channel walls with the variable
porosity screens. In Fig. 3.2.10, the steady-state
temperature distributions for the two screened
systems are compared.

 

6H. F. Poppendiek et al., Analytical and Experimental
Studies of the Temperature Structure Within the ART
Core, ORNL-2198 (Jan. 31, 1957).

’N. D. Greene and W. R. Gambill, ANP Quar. Prog.
Rep. Septr. 30, 1957, ORNL-2387, p 104.

Table 3.2.3. Screen Porosities

 

Porosity

 

Screen Position

Inner Region QOuter Region

 

Collimator 0.580 0.580
No. 1 0.670 0.670
No. 2 0,598 0.665
No. 3 0,557 0.660
No. 4 0.459 0.660

 

QRNL-LR-DWG 30444

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2.0
1.8 // __—.‘-="'-.\
# VARIABLE POROSITY SCREENS
1.6 /./
TER -~
OUTER WALL ‘5? ~
Z 14 / /,/
= UNIFORM
1 / POROSITY SCREENS OUTER WA'-'-\»/
= /
o 4_2 ¥ / ]
o y / |
z A 7
w 40 / // e
2 / FLUID MEAN /// INNER WALL —~ // 7
0.8 ¥ b s 7 /
w / A /Y
Z / V{4
= 086 / 7 7 2
. 7 7 7 L7
/ /= INNER WALL /7 FLUID MEAN
7 N7
04 /L ’
. /[ /‘/ //I
"4
4 L7
0.2 / P s
A 7
o L= é/
o 2 4 6 8 0 12 14 16 80 2 4 6 8 40 42 14 16 18

AXIAL DISTANCE FROM INLET {in.)

Fige 3.2.10. Mean Temperature Distribution for Uniform and Variable Porosity Screened Channels with Two-

Pump Operation. Power generation uniform at 120 kw, NRe = 84,000

78

 

 
The effect of single-pump operation on the
temperature distributions was also studied. The
results are given in Fig. 3.2.11 for the transient
measurements and in Fig. 3.2.12 for the steady-
state measurements. |t is apparent from Fig.
3.2.11 that the screens effectively reduce the
flow asymmetries caused by the one-pump entrance
flow, and hence there is only a small increase
in the temperature fluctuation amplitude between

ORNM 30442

 

 

 

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

¢ o
~ | | I
o 08 VANED-FLOW, OUTER WALL.
5 R
3 "
5 ~
S o061 - N
o r N\
w e — VANED-FLOW, INNER WALL~ TN
o
2 04 —L — -t , ]
< )
§ ] |

/ -
VARIABLE SCREEN, OUTER WALL

& 02 A e ; - 4 ! ——
= k VARIABLE SCREEN, INNER WALL. i

 

 

 

 

 

 

0 | --..-===JF—_1_-:1’;':::-_{:::_3__ == !
8 8 10 12 14 6 i
AXIAL DISTANCE FROM INLET (in.)

Fig. 3.2.11. Transient Temperature Fluctuations for
Vaned-Flow and Variable Porosity Screened Channels
with One-Pump Operation. Power generation uniform at

120 kw; Ng_ = 40,000.

"
ORNL-LR-DWG 30443

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.6
’I
’/
1.4 ~
’/
< ’d’
O 1.2 7 -
= OUTER WALL -] b -
.<_l' { \,’ ,/
E 1.0 ! ,’ /1‘
S INNER WALL ~~_ ] /
|
i 0.8 ’>\ "/ //
I‘i“l ,J > /
E /, 7/ /
é 0.6 ; 77—
w 7 47/ T—MEAN FLUD
= /
w 04 7
A
- / ,,
’ ’
0.2 L e
A A -~
”, -
O ”

 

 

0 2 4 6 8 10 12 14 16 18
AXIAL DISTANCE FROM INLET (in.)

Fige 3.2.12. Mean Temperature Distribution for
Variable Porosity Screened Channel with One-Pump
Operation. Power generation wuniform at 120 kw;

Np . = 40,000+

PERIOD ENDING MARCH 31, 1958

the two-pump and one-pump conditions. This
study has demonstrated that the amplitude of
the surface temperature fluctuations can be
markedly reduced by introducing properly designed
screens into the main channel flow region.

Liquid-Metal Experiment
G. L. Muller® H. W. Hoffman

initial studies of heat transfer in a liquid metal
(mercury) with internal heat generation have been
completed. @ The scope of the experiments is
indicated in Table 3.2.4. To determine whether
magneto-hydrodynamic effects existedin the mercury
flowing through the test section due to the heavy
(1000 amp) electrical current used in heating
the mercury, several half-power runs were made.

Table 3.2.4. Test Conditions for Volume-Heated

Mercury Experiment

 

Run Description

 

16 runs at flow rates ranging
from Np, =6,000 to

Ngp, = 150,000

2 runs at flow rates dupli-

Full power

Half power

cating two full-power tests
Low power 1 run

Effect of guard heater 2 runs at different guard-

on test-section tem- heater power levels

perature

Thermocouple calibration 4 runs at femperature levels
varying fram 58°F to

222°F

Electrode heat loss 2 runs at two power levels

 

The results of these tests can be compared with
the results of full-power runs at the same flow
rates to establish the magnitude of this effect.
Preliminary calculations have indicated that
magneto-hydrodynamic corrections will be negli-
gible. One low-power test was made to establish
the shape of the fluid mixed-mean temperature
profile along the test section.

Midway in the series of full-power experiments,
the steel rotor of the Moyno pump began to bind

 

8On assignment from Pratt & Whitney Aircraft.

79

 
ANP PROJECT PROGRESS REPORT

within the rubber stator. The pump is a positive-
displacement screw-type in which the turning
rotor rubs along the stator, and the liquid being
pumped provides the necessary lubrication between
the rotor and stator. Because of its high surface
tension relative to the usual oxide coating on
steel, mercury is a poor lubricant. In order to
overcome this defect, sufficient sodium metal
was added to the system to produce a sodium-
mercury alloy with 0.02 wt % Na. This sodium
addition had a negligible effect on the thermal
and electrical properties of the fluid, but it
reduced the oxide layer on the steel and allowed
the mercury to wet the steel.

Preliminary analyses of the data indicate that
the surface-to-fluid mean temperature differences

80

are greater than those predicted by the analyses
of Poppendiek.” From this, it appears that the
theory-vs-experiment discrepancies observed in
liquid-metal wall-heat-transfer systems exist
also in the volume-heated system. Since inter-
facial resistances, such as gaseous films or
particulate scales, do not affect the results of
this study (adiabatic walls), it can be concluded
that the radial conductivity (molecular and eddy)
in the mercury is smaller than postulated.

PO .

 

4. F. Poppendiek and L. D. Palmer, Forced Con-
vection Heat Transfer in Pipes with Volume Heat

Sources Within the Fluids, ORNL-1395 (Dec. 17, 1952).
PERIOD ENDING MARCH 31, 1958

3.3. INSTRUMENTATION AND CONTROLS
H. J. Metz

Instrumentation and Controls Division

DATA ACQUISITION SYSTEM
G. H. Burger

A data acquisition, or digital recording, system
was purchased from the G. M. Giannini Company
in order to evaluate the use of such a system in
the handling or recording in digital form of such
variables as temperature, flow, and
pressure which are normally encountered in re-

process

actor instrument systems, reactor component test
stands, and reactor instrument development and
it is expected that, with the experience
gained with the use of this equipment, a better
evaluation can be made of the need for instruments
and instrument systems for future reactor appli-
cation. The system purchased was selected upon
the basis of itsrelative simplicity and compatibility
with the present use of strip-chart recorders, as
well as its expected reliability (comparable with
that of a Brown recorder) and its flexible design
features.

The system is constructed to record automatically
on a typewriter and to display visually on a lamp
bank the values of the millivolt inputs of ten
12-point Brown strip-chart recorders (120 points)
plus a time reference. The automatic typewriter is
equipped to punch tape in the Oracle code. Controls
are included in the system to enable automatic

testing.

logging cycles tobe made at regular timed intervals
in tabular form or on tape or both. The system is
capable of logging 120 points in approximately 4
to 8 min, depending on the Brown recorders. Re-
corders other than the Brown instruments can be
used, and the recorders may be single-point, 16-
point, or 24-point as well as the presently installed
12-point recorders.

A self-balancing potentiometer or an equivalent
device is used to convert voltage inputs to shaft
position. The initial digital encoding is done by
an encoding disk which is mounted on the movable
shaft of the self-balancing potentiometer or other
device,

LIQUID-METAL-LEVEL TRANSDUCERS

G. H. Burger

Life tests of ORNL-designed liquid-metal-level
transducers have been continued. Two of the four

units being tested have accumulated 5408 hr of
operation at 1200°F in NaK and the other two have
accumulated 775 hr. it is planned to continue
these tests indefinitely.

Eight transducers were tested to investigate
wetting at temperatures of 300, 400, 500, and
600°F. Seven of the transducers were regular
Inconel units and one was an Inconel unit plated
with iron to a thickness of approximately 0.003 to
0.005 in. The test results for all the units are
now being compiled and will be included in a
later report.

After completion of the wetting tests, four units
were installed in the test rig and the temperature
was raised to 1200°F for life tests of these units.

STRAIN GAGES FOR USE
AT HIGH TEMPERATURES

C. L. Pearce, Jr.!

A search of the literature on bonded strain
gages for static measurements at temperatures up
to 2000°F is under way. Since the bulk of the
published work on the subject is several years
old and, in many cases, preliminary, some of the
people who have been involved in strain gage
research are being contacted to bring the survey
up to date. Various groups at ORNL who have
used strain gages have been interviewed; however,
the use of strain gages at ORNL has been almost
entirely at low temperatures. The companies that
market commercial strain gages are being contacted
to determine the present limits of applicability of
their gages to high-temperature static strain
measurements and to determine the feasibility of
extending the range.

Effort is being directed toward devising a rela-
tively inexpensive test facility for evaluating and
comparing the various types of bonded strain
gages presently available.  The comparisons
should show wherein the high-temperature limitation
occurs in each case in order to indicate the
possibilities of extension of the temperature limit,

A preliminary survey has indicated that the
present temperature limit on reliable measurements

 

10n assignment from Radio Corp. of America.

81

 
of static strains is 900°F. Phase changes in the
materials appear to be the limiting factor.

TEST SET-UP FOR STUDY OF EFFECT OF
RADIATION ON THERMOCOUPLES

A. M. Leppert

Preliminary design of an in-pile loop in which
to study the effects of high-level radiation on
thermocouples has been completed. In this design,
sheathed thermocouples are brazed into a pot
partially filled with triple-distilled sodium under
high vacuum. The sodium acts as a thermal heat
sink. The pot temperature is obtained by monitor-
ing the vapor pressure of the sodium.

A bench test is under way in order to determine
the reproducibility of the vapor pressure-vs-
temperature curve for sodium. Since the sodium
vapor will be radioactive, it should not be brought
outside the test cell to the electromanometer
pressure transducer, Consequently it was decided
to put an isolating diaphragm in the sodium vapor
line, with a solid-filled system from the diaphragm
to the electromanometer.

A diaphragm and housing have been assembled
and tests begun to determine effects of operating
the filled part of the system and electromanometer
head at varying temperatures and different fill
pressures. A small pot with several sheathed
thermocouples has been fabricated, and as soon
as the diaphragm tests are completed the pot will

" be welded to the diaphragm housing and filled.

82

Tests will be then conducted to check the re-
producibility of the vapor-pressure curve. The
temperature range to be investigated is 600 to

1500°F.

PRESSURE TRANSMITTERS FOR
USE AT HIGH TEMPERATURES

C. M. Burton

Two pneumatic 0- to 200-psi pressure trans-
mitters with 3- to 15-psi signal outputs (Taylor
Instrument Cos., Model 226R)} and Inconel bodies
are now undergoing overrange pressure tests.
These tests were started in an attempt to determine
the aging characteristics of the instruments with
respect to accuracy and to over-all safety. Cali-
brations are made at intervals to detect error
drift, and it is hoped that the errors will give a
warning of impending failure of the primary seal.
It is possible that the primary seal will deform
slightly by bulging of the flat circular plates above
and below the diaphragms. If so, the stresses will
be somewhat relieved as a spherical shape is
approached, creep at the high temperature will
decrease, and the safe life of the unit will be
extended, but it is possible that the accuracy of
the instrument may be affected. The furnaces are
shut down occasionally and micrometer measure-
ments are made on the pressure transmitter bodies
to detect deformation. After approximately 2500
and 1750 hr, respectively, with the two units at
1400°F and 300 psig, calibrations have not changed
appreciably, ond the bodies have not deformed
noticeably.

 
Part 4
SHIELDING

E. P. Blizard
Applied Nuclear Physics Division

 

 
 

 
.

4.1, SHIELDING THEORY

F. L. Keller
Applied Nuclear Physics Division

A CALCULATION OF ENERGY SPECTRUM
AND ANGULAR DISTRIBUTION OF
GAMMA RAYS AT SURFACE OF
BSF REACTOR

L. A. Bowman! R. H. Lessig?
W. W. Dunn? D. K. Trubey

In the past it has not been possible to use
directly the Monte Carlo shielding codes to check,
or to predict, the experimental results of shielding
experiments performed with the Bulk Shielding
Facility (BSF) reactor. While theoretical calcu-
lations of the energy spectrum have been per-
formed3 for comparison with the experimentally
measured gamma-ray spectrum,? there is only
limited information on the angular distribution at
the reactor surface.® The calculation reported
here represents a new attempt to calculate the
energy spectrum from primary gamma-ray sources
and, in addition, to determine the anqular distri-
bution of the gamma rays at the reactor surface.

Geometry

The reactor was divided into four quadrants, with
each quadrant containing one quarter of the re-
actor volume, as shown in Fig. 4.1,1. The thermal-

UNCLASSIFIED
2—0t—059-279

 

 

 

 

//
7
7
7 / \
i’ /.
y SEGOND

QUADRANT

 

 

 

NORTH FACE

 

Fig- 4.‘-1-
the Gomma~Ray Spectrum of the BSF Reactor {Loading
33).

Assumed Geometry for the Calculation of

neutron flux in the second quadrant was used to
perform an integration of the power and the gamma-
ray leakage out the center of the north face of the
reactor. The integration was performed over the
angles ¢ and 6 so that an angular distribution of
the gamma rays could be computed.

Thermal-Neutron Flux

Thermal-neutron flux measurements were avail-
able® for points within the volume of the second
quadrant of BSR loading 33 for a reactor power
of 27.36 kw. Plotting and cross-plotting these
measured values produced curves which could
be used to determine the flux at any point in this
volume. Integration over the reactor, that is,

Power = Cf qb'h(hé)dhé
R

eactor

S L
- sdpd qf’fh IQSI

R? 4R sin 6 d0 d¢ ,

where
¢,,, = thermal-neutron  flux
plots),
C = Ef(u)/K,
Ef(u) = macroscopic uranium fission cross section
= 0,0565 cm™!,
K = factor to convert from fission rate to
power

= 3.48 x 10'3 fissions/sec-kw,
yielded a power of 29,48 kw or approximately 8%
more than the rated power of 27.36 kw. The

(taken from the

 

10n assignment from Wright Air Development Center.
2On assignment from USAF.

3G. deSaussure, A Calculation of the Gamma-Ray
Spectrum of the Bulk Shielding Reactor, ORNL CF.-57-
7-105 (July 31, 1957); ANP Quar. Prog. Rep. Dec. 31,
1956, ORNL.2221, p 343, esp p 347.

4e, C. Maienschein and T. A. Love, Nucleonics

12(5), 6 -8 (1954).

SF. C. Maienschein, T. A. Love, and F. T. Bly,
Gamma Radiation in a Divided Shield. Parts 1 and 2,
ORNL-1714 (Aug. 20, 1954).

6E. B. Johnson, Power Calibration for BSR Loadine
33, ORNL CF-57-11-30 (Nov. 28, 1957).

85

 
ANP PROJECT PROGRESS REPORT

29,48-kw figure was used in the normalization
for this calculation, since neither the integration
over ¢ nor the integration over @ (power and
gamma-ray leakage) accounts for the corners
properly. For the calculation of the second
quadrant to apply equally to the other three quad-
rants, it was necessary to assume that the re-
actor has flux symmetry in all quadrants.

Gamma-Ray Sources

The following primary gamma-ray sources were
considered in this calculation: (1)} prompt fission
and capture gamma rays from uranium, (2) fission-
product-decay gamma rays, (3) capture and decay
gamma rays from aluminum, and (4) capture gamma
rays from water, These sources were grouped
into six initial energy groups of 0.5 1, 2, 4, 6,
and 8 Mev and were taken into account as ex-
pressed in the following equation, which gives
the source for a particular energy group at a
point:

(1) S(EG,R) = by (R) [N(pre )X () + NUfp)E. (u) +
+ N(ADZ (A1) + N(H)Ea(H)] '

where | S
d)fh(R) = thermal-neutron flux at R,
N(p+c) = number of prompt and capture gamma
rays in the energy group about E0
produced in uranium per fission (see

Table 4.1.1),

Ef(u)zmccroscopic fission cross section of
U233 (0.025 ev),
N(fp) = number of fission-product gamma rays
in the energy group about E_ produced
per fission (see Table 4.1.1 and ref 7),
N(Al) = number of aluminum capture and decay
gamma rays in the energy group about
E, per capture (see Table 4.1.1 and
ref 7),
Ea(Al)=mocroscopic absorption cross section
for aluminum (0.025 ev) (see ref 8),
N(H) = number of hydrogen capture gamma
rays in the energy group about E, per
capture (see Table 4.1.1 and ref 7),
3, (H) = macroscopic cross section of hydrogen
per absorption (0.025 ev) (see ref 8).

Absorption Coefficients

The linear absorption coefficients, p, for the
core as a function of E_ were determined by the
following expression, where f; is the fractional
density of the ith material:

@ e (core) = ) G‘) /i _

3) f= Ll

 

7H. W. Bertini et al., Basic Gamma-Ray Data for ART
Heat Deposition Calculations, ORNL-2113, pp 9, 16
(duly 5, 1956).

8H. C. Claiborne and T. B. Fowler, The Calculation
of Gamma Heatinec in Reactors of Rectanguloid Geometry,
ORNL CF-56-7-97, p 20 (July 20, 1956).

Table 4.1.,1. Equivalent Number of Gamma Rays in the Core for Each Energy Group E,

 

 

Energy E N{ptc) N(fp) N(H) N(AI)
Interval (Meg) per per per per
(Mev) Fission* Fission* Capture** Capture**

0 - 0.75 0.5 3.31 2.972 0 0
0.75 -~ 1.5 1.0 2.55 1.864 0 0.10
1.5 - 3.0 2.0 1.687 0.891 1.115 1.53
3.0 - 5.0 4.0 0.368 0.1165 0 0.77
5.0 -7.0 6.0 0.050 0.0084 0 0.21 |
7.0 -9.0 8.0 0.0067 0.0006 0 0.35

 

*Taken from ref 7.

**Taken from ref 8.

 
Buildup Factors

The energy buildup factors for each energy
group were obtained by fitting curves between
the NDA tabulated values’ for aluminum and for
water, Where possible, the curve was weighted
by the ratio of the electron densities of aluminum
and water,

Flux Computation

The uncollided and collided fluxes ina particular
direction at the center of the north face for each
initial energy group were computed by numerically
integrating the standard point kernels over the
volume, V, of the differential solid angle, where
S(EO,I_Q’) is the averuge source strength at [_2’ and
and B(uR) is the energy buildup factor:

(4a) Pyuncoltidedli®:Eq)

f E,S(E,R) e "FR
v

47R?

dR

(4b) qb’)’collide d(e'('b’EO)

E, S(EO,E) e PR .
=f—-— [B(uR) — 11 4R .
v 4nR?

The collided flux from each initial energy group
was separated into effective contributions to the
energy group under consideration and each lower
energy group. For example, for the collided flux
from the 4-Mev initial energy group:

(5) ¢')’collided(E0 = 4 Mev) =q5,),(4 Mev) +
- 42 Mev) + b, (1 Mev) + b, (0.5 Mov) .

The contributions to each of the lower energy
groups were obtained by comparing the areas in
each energy group under the NDA differential
energy spectrum for a point source after passage
through 1 mfp (mean free path) of water for the
initial energy group under consideration.? The
use of the energy spectrum after passage through
1 mfp of water was assumed, since the spectrum
is nearly independent of distance of penetration

 

qH. Goldstein and ). E. Wilkins, Jr., Calculations of
the Penetration of Gamma Rays, NYQ-3075 (June 30,
1954).

 

PERIOD ENDING MARCH 31, 1958

for a number of mean free paths. The total flux
in a particular direction, (;by(B,qb,E), was the sum
of the uncollided flux for that energy group and
the collided contributions from each higher energy
group, This total flux and the uncollided flux
are given in Table 4,1,2. The total and uncollided
flux for @ = 0 was converted to an energy spectrum
and plotted in Fig. 4.1.2 as a histogram. Also

UNCLASSIFIED
2—04—058—0—204R{

PERIMENTAL
7 [CORRECTED VALUES FROM
10 NUCLEONICS 12, 5-6 {1954)]

w

N

o
18]

HIGH -ENERGY
CALCULATION

BY DESAUSSURE,
ORNL-2221, p347

M

o
H

PRESENT CALCULATION

w

HIGH-ENERGY CALCULATION BY
DeSAUSSURE, ORNL CF-57-7-105

T.GAMMA RAY FLUX (phofons/cmz-sec-Mev-w-sfercdian)
o

 

2
103
5
2
10?

o 1 2 3 4 5 6 7 8 9 10 4

£, GAMMA RAY ENERGY (Mev)
Fige 4412, Comparison of Calculated ond Experi-

mental Spectra of Gomma Roys ot the Surface of the
BSF Reactor. Polar angle 0=0 deg.

87
ANP PROJECT PROGRESS REPORT

Table 4.1.2. Gamma-Ray Energy Flux as a Function of the Polar Angle, 6,
the Azimuthal Angle, ¢, and the Average Energy, E

 

Gamma-Ray Energy Flux (Mev/cm?: sec:watt:steradian)

 

 

 

E
(Mev) 0 = 0 deg 0 = 30 deg 6 = 45 deg 8 = 60 deg 6 = 90 deg
Total Uncollided Total Uncollided Total Uncollided Total Uncollided Total Uncollided
P = deg

8.0 2.47x 105 2.38x105 2.24x10% 2.16x105 1.92x10° 1.85x 105 1.58x10° 1.54x10° 140x10° 1.36 x 10°
6.0 3.66x10°5 3.33x105 3.230x10% 2.30x105 2.83x 105 259 x10° 2.34x10° 2.18x 105 2.08x10° 1.95x 10°
40 L5 x10% 1.39x 105 1.43x 106 1.24x10° 1.24x10° 1.09x10° 1.04x10° 9,41x10° 9.31x 105 8.48 x 10°
20 3.43x10% 270 x 10°  3.08x 10° 2,42 x 10° 272x 105 220x 108 2,37 x 10% 1.98x10% 215x10® 1.83 x 10°
1.0 2.52x 105 1.38x 106 2.26x10® 1.25x10® 2.00x10% 1,17x10° 1.75x10® 1,10x10® 1.60x10° 1.04 x 10
0.5 451 x 105 6.69x 105 3.98x 105 6.20x 105  3.46 x 105 5.94x 105 2.93x10® 575x 105 2.63x 10% 551 x 10°

¢-30 deg

8.0 247x105 2.38x 105 2.12x10° 204x105 1.92x10% 1.86x10° 1.64x105 1.5 x 105 1.34x10% 1.31x10°
| 6.0 3.66x10% 7333105 3.12x 105 2.85x 105 2.84x 105 2,60 x 105 2.43x10° 2.25x10° 1,99 x 10° 1.8 x 10°
| 40  1.59x10° 1.39x10°5 1.36x10% LI9x10® 1.24x10® 1,10x10® 1.08x10% 9.66x10° 8.96x10° 8.18x10°

2.0  3.43x108 270x 108 2.90x106 237x10% 2.73x10® 219x10® 2.43x10° 201x10% 2.08x10% 1.77 x 10

1.0 2.52x105 1.38x 108  2.10x10° 1.24x10® 2.01x105 1L18x10® 179x10® 1.11x10% 1.54x10® 1.02x 10°

0.5 451 x 105 6,69 x 105 3.70x 105 620 x 105  3.46 x 10° 598 x 105 3.01x 10% 578x10% 2,50x10% 545x 103
|
|
\

¢ = 50 deg
8.0 2.47 x 105 2.38x 105 2.08x 105 2.01x105 1.76x105 1.71x105 1.46x10° 1.43x10° 119x10% 1,17 x 10°
6.0 3.66x10% 3.33x105 3.06x 105 2.81x105 2.60x105 241x105 217x10° 2.04x105 177x105 168 x 10°
4.0 159 x 105 1.39x 105  1.34x10® 1.18x10° 115x10% 1.04x10® 9.75x10° 8.88x 105 8.04x10° 7.45x10°
2.0  3.43x 105 270 x 105 2.94x 105 236 x 105  2.60 x 106 2,14x 10® 226 x10° 1,92x10% 191 x10® 166 x 10°
1.0 2.52x105 1.38x10% 216 x 105 1.24x10% 1.92x10® 1.17x 106 1.62x10® 1,08x10% 1.42x10% 9.83 x 10°
0.5 451 x 105 6.69x 105 3.78x10° 6.23x 105 3.27x10% 598x105 276x10% 571x105 225x10% 537 x 10°

¢ = 65 deg
8.0 2.47x 105 238x10° 205x 105 1.98x105 1.65x105 1.61x10% 136x10% 1.33x10° 112x10° 110x10°
6.0 3.66x105 333x105 3.03x10° 278x105 2.45x105 229x10° 2.02x10° 191x10° 1.66x10° 1.58x10°
40  1.59x105 1.39x10% 1.33x106 1.18x 105 1.09 x 10¢ 9.89x 105  9.16 x 10° 8,42 105 7,57 x 10°  7.07 x 10°
2.0 3.43x 105 270 x 105 2,96 x 108 2.41x 105  2.50 x 106 2.09 x 10¢ 2.15x 10® 1.85x 10%® 1.B2x 10¢ 161 x 10
1.0 2.52x10° 1.38x 105  2.15x10% 1.25x 10 1.85x10° 1,16 x10® 1,59 x 108 1.07 x 10 136 = 10®  9.64 x 10°
0.5 4.51 x 10°  6.69 x 105 3.75x 105 6.28x 105  3.12x 105 6.00x 105 2,59 x 106 571 x 105 2,09 =105 534 x 10°

¢ = 90 deg
8.0 2.47x10% 2.38x 105 2.05x105 1.98x10° 1.62x10% 1.58x10% 1,33x10° 1,30x10° 1.08x10° 107 x 10°
60 3.66x 105 3.33x105 3.03x10° 279x10% 2.39x105 2.25x10° 1.97x10° 1,87 x10° 161 x10° 1.55x10°
40  1.59x10° 1.39x108 1.33x 105 1.1Bx10C 1.07 x 104 977 x 105 8.93x10% 825x10° 7.40x10° 6.93 x 10°
20 3.43x10° 270x 105 2.96x 10 239 x 10° 3.47 x 105 2.08x 105 2,11 x10° 1.83x10® 1.79x10® 1.59x 10°
10 2.52x105 1.38x 105 2,17 x10° 1.27x 105 1.83x10% 116x 105 1.56x10% 1.06x10® 133x10% 9.60 x 10°
0.5 450 x105 6.69x 105 3.80x10° 6.35x10% 3.08x10% 6.05x10° 252x10% 573x10° 203x10® 535x 10°

 

88

 
PERIOD ENDING MARCH 31, 1958

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

shown are the results of calculations by Conclusions
3 . 4
deSaussure” and the experimental results, The gamma-ray energy spectrum normal to the
Anaulos Distributi reactor surface, as calculated in this study, is in
ngular Distribution good agreement with previous calculations by

The energy flux at the surface-of the reactor, deSaussure.® The present calculation is an over-
averaged over the azimuthal angle, ¢, is shown estimate, particularly in the low-energy range,
in Fig. 4.1.3 as a function of the polar angle, 6. since it was assumed that the collided flux, as
A typical plot of the variation of the flux with ¢ calculated by the energy buildup factors, is all
when 6 = 30 deg is shown in Fig. 4.1.4. traveling in the outward direction,

UNCL ASSIFIED 6 NCLA FIED

"‘5’06) 2-01-059 - 28 (X1g ) 2201~ 089 - 580

£=0.5 Mev !
\\ =
4 - 5 4
T |~ £ =0.5Me
' ‘é \________“ — ° :

.E £ =2 Mey : ..'5
E 1’\\4 \\ ;

2 o

v \ .4 F =2 Mev

_3 3 k N.E 3 M—— S . ¢
mt'n AE=1 Mey \ \\ :>:

% = |y______-.- £ =1 Mev
2 2 T~ Lfi- 2 e — e}
iy I ‘ <
D - ) o
N — 8

——— S ! £ =4 Mev
\\ Il °
@
1 [ o
] <
£ =6 Mey £ =6 Mev
e : I—t—
o £=BlMev *T\ $ — o £ =8 Mev
0 ‘6 20 30 40 50 60 70 80 90 0 10 20 30 40 50 60 70 80 90
8, POLAR ANGLE (deq) ¢, AZIMUTHAL ANGLE (deg)

Fige 4+1:3. Total Calculated Gamma=Ray Energy Fige 4.1.4. Total Calculated Gamma-Ray Energy
Flux at the Surface of the BSF Reactor as a Function Flux at the Surface of the BSF Reactor as a Function of
of the Polar Angle, 6, Averaged over the Azimuthal the Azimuthal Angle, ¢, for Various Energies E When
Angle, ¢, for Various Energies E. Pclar Angle 0 = 30 deg.

89

 

 
ANP PROJECT PROGRESS REPORT

4.2. LID TANK SHIELDING FACILITY

W. Zobel
Applied Nuclear Physics Division

STUDY OF ADVANCED SHIELDING
MATERIALS (GE SERIES)

L. Jung

The experiment designed by GE-ANP for studying
the production of secondary gamma rays in con-
figurations containing advanced shielding materials
is being continued. The configurations used in
the latest tests are listed in Table 4.2,1, and the
size and composition of the various materials in-
cluded in the configurations are described in
Table 4.2.2.

these configurations have been completed, only the

While all the measurements beyond

gamma-ray and fast-neutron measurements are
reported here; the thermal-neutron data have not
yet been evaluated.

The group 2 configurations consisted of a 4-in,-
thick beryllium slab, 4 in. of stainless steel as a
gamma-ray shield, and a 12-in.-thick lithium
hydride neutron shield with boral placed at various
positions within the configurations in an attempt to
reduce the number of secondary gamma rays pro-
duced in the components of the shield, especially
in the stainless steel. There was an unavoidable
J-cm-thick oil layer adjacent to the source plate.

The effect of inserting boral between the beryl-
lium and the stainless steel in this group of con-
figurations is shown in Fig. 4.2.1. A V-in. thick-
ness of boral (configuration 2-Q) reduced the
gamma-ray dose rate 100 c¢cm from the source by
34.4% of that measured in the case where no boral
was used (configuration 2-A). Increasing the boral
thickness to I/2 in. (configuration 2-F) reduced the
dose an additional 12.2%. A further increase in the
boral thickness to 1 in. (configuration 2-0) gave no
additional attenuation.

A comparison of measurements beyond con-
figurations 2-G, 2-N, 2-0, and 2-P (Fig. 4.2.2)
shows that placing all the boral immediately be-
hind the beryllium was not as effective as dis-
tributing it throughout the stainless steel, as long
as some of the boral remained behind the beryllium.
Placing 1 in. of boral immediately behind the
beryllium (configuration 2-0) reduced the gamma-
ray dose rate 100 cm from the source 17% below
that measured for a configuration in which the
boral was distributed throughout the stainless
steel (configuration 2-P). However, placing '/2 in.
of the boral in the stainless steel and keeping

90

another ]/2 in. behind the beryllium (configuration
2-N) reduced the dose rate at the same point by
27%.

Fast-neutron dose rates beyond configurations
2-F, 2-G, 2-N, 2.0, 2-P, and 2-Q are shown in
Figs. 4.2.3 and 4,2,4. No particular significance
can be attached to the noticeable differences in
the curves, since the reproducibility of the fast-
neutron data at the LTSF is not better than 10%.

Configurations 4.0 and 4-P consisted of 26 ¢m of
oil adjacent to the source plate (not including the
unavoidable l-cm oil gap) followed by 4 in. of
stainless steel. Configuration 4-0 alsc had l/2 in.
of boral immediately behind the stainless steel.
As indicated by Fig. 4.2.5, there were no signifi-
cant differences in the dose rates beyond these
two configurations; however, a comparison of these
measurements with the measurements beyond
configuration 4-A, in which the 26-cm oil gap was
not present, shows that the oil gap in configurations
4-0 and 4-P reduced the gamma-ray dose rate 100
cm from the source plate to 22% of that measured
behind configuration 4-A. The fast-neutron dose
rates (Fig. 4.2.6) showed no appreciable difference
except immediately behind the stainless steel.
The dose rate beyond configuration 4-P is higher
by 45% than that beyond configuration 4-A at a
point 44 cm from the source. At 62 cm the two
curves infersect. The curves for configurations
4-A and 4-0 merge for all distances from the source
>60 cm.

Configurations 4-M, 4-N, 4-Q, 4-R, 4-S, and 4-T
consisted of 4 in. of stainless steel and ]/4 to 1 in.
of boral placed in various positions with respect
Configurations 4-M and
4-N were run to determine the comparative effects
of placing }/2 in. of boral in front of or behind the
stainless steel. The results (Fig. 4.2.7) show
that an additional reduction of 14% was obtained
100 cm from the source when the boral was placed
behind the stainless steel.
dose

to the stainless steel.

A comparison of the
rates beyond configurations 4-N and 4-T
shows that substituting l/4 in. of boral for ]/2 in.
of boral behind the beryllium increased the gamma-
ray dose rate by 13% at a distance 100 cm from
the source.

The measurements beyond group 4 configurations
again showed that evenly distributing the boral
in the stainless steel matrix, while at the same

 

 
 

 

 

PERIOD ENDING MARCH 31, 1958
Table 4.2.1. Description of Latest Configurations Tested in GE Series?
Location of Unavoidable
: Configuration L . . Probe,b Oil Gap Within
Description of Configuration c
Number z, Configuration
- (em) (cm)
2-A 4 in. of Be + 4 in. of stainless steel + 12 in. of LiH + oil 58.0 2.3
2-F 4 in. of Be + l/2 in. of boral + 4 in. of stainless steel + 59.8 2.9
12 in. of LiH + oil
2-G 4 in. of Be + ]/2 in. of boral + 4 in. of stainless steel + 61.0 2.8
]/2 in. of boral + 12 in. of LiH + oi!
2-N 4 in., of Be + ]/2 in. of boral + 2 in, of stainless steel + ]/2 61,8 3.6
in. of boral + 2 in. of stainless steel + 12 in. of LiH +
oil
20 4 in, of Be + 1 in. of boral + 4 in, of stainless steel + 12 61.8 3.7
in. of LiH + oil
2-P 4 in. of Be + 1 in. of sfalniess steel + /2 in. of boral + 2 62.0 3.9
in. of stainless steel + /2 in. of boral + 1 in. of
stainless steel + 12 in. of LiH + oil
2-Q 4 in. of Be + J, in. of boral + 4 in. of stainless steel + 12 59.4 3.3
in. of LiH + oil
4-A 4 in. of stainless steel + oil 15.8 1.0
4-p4 l/2 in. of boral + 4 in. of stainless steel + oil 17.8 3.7
4-N 4 in. of stainless steel+ ), in. boral + oil 17.6 3.5
) 4-0 26 cm of oil + 4 in. of stainless steel + ]/2 in. of boral + oil 42.6 2.0
4-P 26 cm of oil + 4 in. of stainless steel + oil 41.6 2.3
4-Q ]/2 in. of boral + 4 in, of stainless steel + ]/2 in. of boral + 19.6 3.7
oil
4-R 3 in. of stamless steel + / in. of boral + 1 in. of stainless 19.8 3.9
steel + /2 in. of boral + osl
4-5 2 in. of stamless steel + /2 in. of boral + 2 in. of stainless 19,6 3.7
steel + / in. of boral + oil
4-T 4 in, of stainless steel + ]/4 in. of boral + oil 17.6 3.7
41-A 4 in. of stainless steel (dry) + oil in an Al tank (]/s-in.-thick 15.5 1.8 (air gap}
walls)
41-B 4 in. of stainless steel (dry) + ]/2 in. of boral (dry) + oil in an 16.8 1.9 (air gap)
Al tank (¥ -in.-thick walls)
41-C 2 in. of stainless steel (dry)+ / in. of boral (dry)+ 2 in. of 18.0 1.7 (air gap)

sfumless steel (dry) + /2 in. of boral (dry) + oil in an Al
tank (/ in.~thick walls)

Oil Oil only in configuration tank 3.]5d

 

- ?Includes some configurations reported in ANP Quar, Prog. Rep. Dec. 31, 1957, ORNL-2440, p 263.
Actual distance from source plate to back of solid configuration.
“Includes 1-cm recession of the aluminum window in the configuration tank.

dPosifion of inside surface of the aluminum window in the configuration tank.

91
ANP PROJECT PROGRESS REPORT

time keeping a good thermal-neutron absorber
behind the solid materials, is the most effective
way to reduce the capture gamma-ray dose rate.
A comparison of the dose rates beyond configuration
4-A, in which no boral was used, and 4-5, in which
boral was evenly distributed in the stainless steel,
shows that the boral reduced the dose rate 100 cm
from the source by 51% (Fig. 4,2.8), whereas the
dose rate beyond configurations 4-Q and 4-R, in
which boral was used but was not placed within
the stainless steel matrix, was reduced 47%. As
was expected, there were no significant changes
in the fast-neutron dose rates beyond the group 4
configurations (Fig. 4.2.9).

Configurations 41-A, 41-B, and 41-C consisted
of 4 in. of stainless steel interspersed with boral
in a dry arrangement to eliminate oil gaps between
the slabs.
aluminum

The oil was contained in a 56-in.-long
tank which was placed
behind the solid materials. Gamma-ray dose-rate
measurements beyond these configurations (Fig.
4,2.10) show that the addition of ]/2 in. of boral
behind the stainless steel reduced the dose rate
approximately 46% at 100 cm from the source. A

immediately

further 5%reduction was realized when an additional

Y in. of boral was placed within the stainless

steel, but this may be within experimental error.

Comparing measurements beyond configuration
41-C with those beyond a similar configuration,
4-S (in Fig. 4.2.8), shows that the dose rate 30 cm .

behind configuration 41-C was approximately 61%
higher than that beyond configuration 4-5. Since
configuration 41-C did not have the 3.7 cm of
oil distributed throughout the configuration, it is
presumed that the difference in the dose rates is
due to the additional inelastic scattering in con-
figuration 41-C,

The fast-neutron dose-rate measurements (Fig.
4.2,11) show an average reduction of 17% from
configuration 41-C to 41-A caused by the addition
of 1 in. of oil in place of the boral. The fast-
neutron dose rate beyond configuration 4-S is
approximately 35% lower than that beyond con-
figuration 41-C; the comparison is based on equal
oil thickness behind each configuration. This
difference is due to the distributed oil in con-
figuration 4-S.

This series of experiments is being continued.

Table 4.2.2. Physical Properties of Materials Used for Mockup Tests -
of Advanced Shielding Materials (GE Series)

 

Thickness of

 

Material Material* per Slab Description
(cm}

Transformer oil Specific gravity, 0.88 at 15.6°C; nominal composition,
86.5 wt % C and 12.5 wt % H; H:C atom ratio, 1.72:1

Beryllium 10.16 48 X 49 x 4 in, metallic slabs

Type 347 stainless steel 2.54 5ftx5ftx1in.slab; composition, 58.35 wt % Fe,
18.0 wt % Cr, 10.3 wt % Ni, 2.0 wt % Mn, 1.0 wt %
Si, 0.08 wt % C, 0.04 wt % P, 0.03 wt % §; density,
7.8 g/cm:‘1

Lithium hydride 32,0 5x 5x 1ft slab encased in aluminum cans (k‘-in.-thick
walls); composition, 95 wt % LiH and 5 wt % un-
accounted for (probably oxygen and hydrogen);
density, 0.75 g/cm3

Boral 1.3 48 X 48 X 0.4 in. slab with 0.05-in.-thick Al clodding;
composition, 80 wt % Al, 15.7 wt % boron, 4.3 wt %
carbon; density, 2.60 g/cm3

Boral 0.63 44,5 X 44,5 < 0.15 in. with 0.05-in.-thick Al cladding;

composition, 80 wt % Al, 15.7 wt % boron, 4.3 wt %
carbon; density, 2.60 g/Cm3

 

*Including cladding.

 

 
PERIOD ENDING MARCH 31, 1958

QI
2—04—057— 71— 444

BORAL S.S.
(in.) (in.)

S 4

4
4
4

N

S

GAMMA -RAY TISSUE DOSE RATE (ergs/g-hr'watt)
o

N

 

5
g x 10"
. T0 30 110 130 {50 170
Z4 s DISTANCE FROM SOURCE PLATE {cm)
¢ Fig. 4.2.1. Gommo-Ray Tissue Dose Rates Beyond Configurations 2-A, 2-F, 2-0, and 2-Q: Effect of Inserting

Boral Behind Beryllium in Configurations Containing Stainless Steel.

 
ANP PROJECT PROGRESS REPORT

S
2—01—057—71—4145

RS
4 %X 10

N\ 4

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr- watt)

5
Be BORAL S.S. BORAL BORAL S.S.
(in.) (in,) (in.) (in.) in.) (in.)
4 4 o
4 2 Y

> 4 4
4 1 Yo

 

10
70 a0 140 130 150 170 190 .
Z,, DISTANCE FROM SOURCE PLATE (cm)

Fig: 442.2. Gamma-Ray Tissue Dose Rates Beyond Configurations 2-G, 2-N, 2-0, and 2-P: Effect of Distributing
Boral Throughout Configurations Containing Stainless Steel.

94

 
FAST-NEUTRON TISSUE DOSE RATE (ergs/g-hr+ watt)

 

PERIOD ENDING MARCH 31, 1958

Ay s
1 2-01-057-71-416

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

A
\é\o
CONF. Be BORAL S.S. BORAL S.S. BORAL S.S. LiH
NO. {in.) (in.) {in.) {in.) {in.} (in.) (in.) {in.)
o 26 4 Vo 4 7 12 +0iL
2 A 2N 4 Yo 2 Yo 2 12 +O0IL |
® 20 4 1 4 12 +0IL
a  2-p 4 { V2 2 Yo 1 12 +0IL
16° -
65 70 75 80 85 90 95 100

Zy» DISTANCE FROM SOURCE PLATE {cm)

Fig. 4.2.3. Fast-Neutron Tissue Dose Rates Beyond Configurations 2-G, 2-N, 2.0, and 2-P,

95
ANP PROJECT PROGRESS REPORT

-1 2—01—-057-71-417
10 '

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FAST-NEUTRON Ti{SSUE DOSE RATE (ergs/g-hr-wott)
o

 

 

 

 

 

 

 

 

 

 

 

 

 

B
CONF. Be BORAL S.S. LiH
2 NO. (in.}) (in.} (in.) {in)
o 2-F 4 o 4 12 +0IL
e 2-Q 4 Va 4 12+ OIL
1073 - .
65 70 75 80 85 90 95 §00

Zgs DISTANCE FROM SOURCE PLATE (cm)

Fig. 4.2.4. FastNeutron Tissue Dose Rates Beyond Configurations 2-F ond 2-Q.

 
GAMMA - RAY TISSUE DOSE RATE (ergs/g-hr-wcH)

OlL
(cm)

S.S.
(in.)

4 +0IL

BORAL
(in.)

+0IL
+0IL

26 4 2
26 4

50 70 90

ZO‘

{10

PERIOD ENDING MARCH 31, 1958

R
2-01-~057-T7i-418

 

130 150

DISTANCE FROM SQOURCE PLATE {(cm)

Fige 4.2.5. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-0, and 4-P:

Preceding Configurations Containing Stainless Steel.

Effect of Oil Gap

97

 
ANP PROJECT PROGRESS REPORT

-t - L]

2-01-057-71—-419

FAST—NEUTRON TISSUE DOSE RATE {ergs/g-hr-watt)

CONF. BORAL
NO. {in.)

O 4-A +0IL
® 4-0 iz +0IL
A 4-pP +0IL

 

40 50 60 70 80 90 {CO {10
2y, DISTANCE FROM SOURCE PLATE (cm)

Fig. 4¢2.6. Fast-Neutron Tissue Dose Rates Beyond Configurations 4-A, 4-0, and 4-P. .

 
10°

GAMMA -RAY TISSUE DOSE RATE (ergs/g-hr- watt)

20 40 60

BORAL

80

{in.)

1 /2

 

PERIOD ENDING MARCH 31, 1958

_
2—04—057—-71-420

S.S. BORAL
(in.) {in.)

4 +0IL
4 +0IL
4 Y. +oIL
4

1%, +oIL

100 120 140 160

z,,DISTANCE FROM SOURCE PLATE (cm)

Fig. 4.2.7. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-M, 4-N, and 4-T: Effect of Inserting

Boral in Configurations Containing Stainless Steel.

 

99

 
ANP PROJECT PROGRESS REPORT

<
2—01—-057—-71— 424

GAMMA—RAY TISSUE DOSE RATE (ergs /g-* hr-watt)

BORAL S.S. BCRAL S$.S. BORAL
(in.) (in.) (in.) {in.) (in.)

a \
Yo 4
3 s
2 ‘/2 \

20 40 60 80 100 120 140 160 -
DISTANCE FROM SOURCE PLATE (cm)

 

Za

Fig. 4.2.8. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-Q, 4-R, aond 4-S: Effect of Inserting *
Boral in Configurations Containing Stainless Steel.

100

 

 
 

PERIOD ENDING MARCH 31, 1958

A
2-M-057-T1-422

CONF. BORAL S.S. BORAL S.S.
{in)) {in.) {in.) {in.)

1/2

FAST-NEUTRON TISSUE DOSE RATE (ergs/g-hr-watt)

 

20 30 40 50 60 70 80 90 100 HO 120 130 140
ZO,D|STANCE FROM SOURCE PLATE SURFACE (cm)

Fig. 4.2.9. Fast-Neutron Tissue Dose Rates Beyond Configurations 4-A, 4-Q, 4-R, 4-S5, and 4-T.

101

 
| ANP PROJECT PROGRESS REPORT

e o
4 2-0t—057 —71-423

 

CONF S.S BORAL S.S. BORAL .
NO. (in) (in.} (in.) {in.)

41-A + OIL iN Al TANK
41-8 + OIL IN Al TANK
a1-C Y2 +0IL IN Al TANK

NOTE: ALL SOLID COMPONENTS KEPT DRY

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr-watt)

20 40 60 80 - {00 120 140
Z5, DISTANCE FROM SOURCE PLATE {(cm)

Fig. 4:2.10. Gamma-Ray Tissue Dose Rates Beyond Configurations 41-A, 41-B, and 41-C: Effect of Inserting
Boral in Configurations Containing Stainless Steel.

102

 
PERIOD ENDING MARCH 31, 1958

| <
2-01-057-T1-424

FAST—NEUTRON TISSUE DOSE RATE (ergs /g - hr - watt)

 

107"
5
CONF. 5.5. BORAL BORAL
NO. (in) (in.) (in.)
2 41-A 4 + OIL IN Al TANK
4-B 4 Yo + OIL IN Al TANK
10-2 M-C 2 . Yo + OILIN Al TANK
NOTE: ALL SOLID COMPONENTS KEPT DRY
5
2
=3
10
. 20 30 40 50 60 70 80 90 100

Zy, DISTANCE FROM SOURCE PLATE (cm)

Fig. 4:2.11. Fast-Neutron Tissue Dose Rates Beyond Configurations 41-A, 41-B, and 41-C.

103

 
ANP PROJECT PROGRESS REPORT

4.3. BULK SHIELDING FACILITY

F. C. Maienschein
Applied Nuclear Physics Division

THE BSF STAINLESS STEEL-UO,
REACTOR (BSR-II)

E. G. Silver

As previously reported,' a stainless-steel-clad
UO,-fueled reactor for the BSF is being designed.
This reactor, identified as the Bulk Shielding Re-
actor |l (BSR-Il), is to be used alternately with
the present aluminum-clad reactor (BSR-I).

The utility of the BSR-II for shield test purposes
derives from the nature of its gamma-ray spectrum
and its geometrical configuration. A steel core
construction will give rise to a gamma-ray spectrum
which is harder than that from an aluminum core.
Since the reactors of interest to the ANP program
have, in many cases, core components of the
Fe-Ni-Co system, such core materials in the
test reactor will mock up the gamma-ray spectrum
more successfully than an aluminum core would.

The BSR-ll is designed for maximum utility as
a shield-test radiation source. The core is
cubical, 15 in. on each edge, and the grid plate
is designed so that it does not project laterally
beyond the core edge. The over-all height of the
core is also minimized in order to make compact
wraparound shield mockups possible. This arrange-
ment will provide an intense, symmetrical, localized
source of reactor radiation,

Detailed descriptions of the design of the BSR-I|
are included in the safeguard report for the re-
actor.2 The reactor will utilize a control system
of four spring-driven pairs of control plates,
which can be positioned in any of the 25 fuel-
element positions of the core. The acceleration
of the control plates by means of springs will
assure safety of the system, which, owing to
its high absorption cross section, has a prompt-
neutron-generation time in the core that is only
about one-third as long as that of the BSR-l.
Mechanical tests and analog computations carried
out on the control system indicate that it is

 

1. s. Lewin, ANP (uar. Prog. Rep. March 31, 1957,
ORNL-2274 (Part 6), p 34.

E. G. Silver and 1. S. Lewin, Safeguard Report for a

Stainless Steel Research Reactor for the BSF (BSR-II),
ORNL-2470 {(March 12, 1958).

104

capable of safely reversing positive periods as
short as 8 msec. This would be equivalent to
an excess reactivity of 1,2%. However, as a
safety precaution it is planned to limit the
potential excess reactivity to 0.75% (prompt
criticality) until a planned series of tests on
the BSR-Hl and its control system at the SPERT
Facility are completed.

The safeguard report also contains the results
of calculations carried out with respect to hazards
arising outside the X-10 area from a catastrophe
to the BSR-Il. It was found that the hazards were
less than those which could be caused by the
ORR, which adjoins the BSF, and which has
been approved for operation. It is anticipated
that, following approval by the Advisory Committee
on Reactor Safeguards, the BSR-I| will be fabricated
and placed in operation during the calendar

year 1958,

SPECTRA OF GAMMA RADIATION ASSOCIATED
WIiTH THE FISSION OF u23s

T. A, Love R. W. Peelle
F. C. Maienschein W. Zobel

For some time efforts have been under way at
the Bulk Shielding Facility to measure the spectra
of the gamma radiation which emanates from
U235 (ref 3) and its products® as a result of
fission induced by thermal neutrons. Information
on the resulting gamma-ray distributions, given
as a function of the time after fission and the
photon energy, is essential for optimum design
of both shielding ond heat-deposition systems
for power reactors. Two experimenfcl approaches
have been used thus far: in one approach, delayed-
gamma-ray data were obtained by observing u23s
samples that had been exposed in a reactor; in
the other approach, both delayed- and prompt-
gamma-ray data were obtained by observing spectra
in time-correlation with pulses from a U239 _con-
taining pulse ion chamber. The progress with

 

3F. C. Maienschein et al., Pbys. Semiann. Prog. Rep.
March 20, 1955, ORNL-1879, p 51.

R, w. Peelle, T. A. Love, and F. C. Maienschein,
A[\2{g3 Quar. Prog. Rep. June 10, 1955, ORNL-1896,
P .

 

 
 

 

 

 

 

the experiments and the analyses for these two
types of problems is discussed below.

Fission-Product Gamma Rays Emitted More
Than 1 sec After Fission

Previous reports® have presented the results
obtained from the preliminary analysis of a series
of experiments in which studies were made of
time and energy distributions of gamma rays given
off by irradiated U?3% samples between 1 sec
and 1500 sec after fission. The final analysis
has been delayed by the need to develop methods®
for removing from the data all nonuniqueness
effects caused by the response characteristics of
the multiple-crystal spectrometer used in the ex-
periments. Such corrections to the previously
reported values may amount to between 10 and
30%. Progress has been made, and during the
coming report period it is onticipated that final
analysis will be completed.

The data have been partially reanalyzed to
provide a better approximation to the final result.
The steps which have been completed include
improvements in the calculation of the number of
fissions occurring in the samples used and removal
from the observed pulse-height spectra of the
contributions from the ‘‘tail’’ of the detector
response functions for monoenergetic gamma rays.
Effects from the breadth of the peaks of the
response curves have not been removed but are felt
to be less important. The results of this latest
reanalysis are presented in Figs. 4.3.1 and 4.3.2.
Included in Fig. 4.3.1 is the previously reported’
equilibrium spectrum from a rotating fuel belt,
which has not been reanalyzed. Included in Fig.
4,.3.2 is a comparison with radiochemical data®
of the total observed rate of photon energy release
as a function of time after fission.

 

5W. Zobel, T. A, Love,and R. W. Peelle, ANP Quar
Prog. Rep. March 10, 1956, ORNL-2061, p 250-353;
W. Zobel and T. A. Love, ANP Quar. Prog. Rep. Sept.
10, 1956, ORNL-2157, p 291-96; Appl. Nuclear Phys.
Ann. Rep. Sept. 10, 1956, ORNL-2081, p 95-101.

Sw. Zobel et al., ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 294-95; Appl. Nuclear Phys. Ann.
Rep. Sept. 1, 1957, ORNL-2389, p 97-98.

7R. W. Peelle, W. Zobel, and T. A. Love, ANP Quar.
Prog. Rep. Dec. 10, 1955, ORNL-2012, p 223-26; Appl.
Nuclear Phys. Ann. Rep. Sept 10, 1956, ORNL-2081,
p 91-94.,

J. F. Perkins and R. W. King, ‘“Energy Release from
Decay of the Fission Products,’”” Nuclear Sci. and Eng.
(to be published); Papers Presented at the Fourth
Semiannual ANP Shielding Information Meeting, Nov.
19-20, 1957, ORNL-2497, vol Il, paper II-B.

ated gamma

PERIOD ENDING MARCH 31, 1958

Fission-Product Gamma Rays Emitted Between
5 x 10~8 and 10~% sec After Fission

During the preliminary experiments designed to
aid in the determination of the energy spectrum
of gamma rays emitted in coincidence with fission,
counts were observed® in the region between
5 x 10~8 and 104 sec after fission. These counts
are now believed to arise from delayed gamma
rays. More recently, a beam of thermal neutrons
from the ORNL Graphite Reactor has been utilized
to obtain measurements of the intensity and energy
of these gamma rays. For experimental reasons
that involve random backgrounds, it was impossible
to use the multiple-crystal spectrometer for these
measurements. Instead, a single-crystal sodium
iodide detector was vused; consequently, the
results are presented in terms of pulse-height
spectra rather than energy spectra.

Some of the spectral data obtained at various

short times oafter fission, by using standard
delayed-coincidence techniques, are shown in
Fig. 4.3.3. The source of the fission radiation

0.5-cm-dia by 1.0-cm-long spiral
fission chamber. Time-distribution data obtained
for various energy groups with the same fission
source and the previously described'® time
analyzer are shown in Fig. 4.3.4. Gamma rays
appearing ‘‘after’” fission were compared with
those ‘‘before’’ fission to avoid any effects from
gamma rays coincident with fission. Absolute
intensities of the gamma radiation in these two
experiments are known, approximately, in terms
of the intensity of thé prompt gamma radiation as
observed in the same geometry and plotted in
Fig. 4.3.3. (The plotted points in this case were
obtained by multiplying the observed number of
counts per fission per Mev by 107.) Values for
the relative number of counts observed in a few
pulse-height groups during the time interval after
fission between 5 x 10=8 and 10~% sec are given
in Table 4.3.1.

Although this contribution to the fission-associ-
radiation does not appear to be
compared with the prompt radiation,
short-delayed gamma rays are not known

was a small

negligible
such

 

9F. C. Maienschein, R. W. Peelle, and T. A. Love,

Appl. Nuclear Pbys. Ann, Rep. Sept. 1, 1957, ORNL-
25759, p 99~110. b ¢

IOR. W. Peelle, F. C. Maienschein, and T. A, Love,
Appl. Nuclear Pbys. Ann. Rep. Sept. 1, 1957, ORNL-
2389, p 263-69.

105
 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED UNCLASSIFIED
2-01-058-0-410 y I

 

3 2-01-058-0-4{!
5x10 ) 10
2 5
107" ) . : |
SPECTRUM FROM CONTINUCUS BELT
{USE SCALE AT RIGHT —=)
5 10—1
5
2 T
o
w
1072 2 -
s
-2 @
5 10 +
T
5 g
2 @
o
» 2
T -3
', 10 o
3 T -3
T' Q 10
L T
g g 5 J—
.F @ - TQOsec
T2 £ =
& T
= -
g 10 1000 sec o
= E 10_4 e
L Q
£ 5 £ . 1550sec
5
2
2
107°
107"
5
5
2
Z
e COMPTON —|— PAIR GOMPTON —|— PAIR —=
10 10-6
S 5
2 2
o’ 1077
O 1.0 2.0 3.0 4.0 50 6.0 O 1.0 2.0 3.0 4.0 5.0 6.0
ENERGY {Mev) ENERGY (Mev)

Fig. 4.3.1. Gamma-Ray Energy Spectra Observed at Long Times After Fission. The errors shown are a combination of counting statistics with other uncertainties. The lines were fitted to the points by eye. All
data, except for the top curve in (b), were obtained at the stated times after fission. The top curve in () is the spectrum emitted from the continuous fuel belt about 8 x 103 sec after irradiation of the belt commenced
(use righthand scale). No response corrections were included in this top curve.

106

 

 
UNCLASSIFIED
ORNL-LR-DWG 20558A

x10

-1

 

T3 102 > 8
= T
- c
. k)
? Mev @
@ &
T O5i-112Mev| ~ _, 3
§10° t 2
owr
o
&
o
c
o
2
Syg -5
10°% 6
10®
1 2 5 10 20 50 100 1000 10,000 50000
TIME AFTER FISSION {sec)
Fig. 4.3.2, Time Dependence of the Emission of

Gomma Rays for Long Times After Fission.

lower

On the
graph are shown the photon

intensity-time distributions for six energy intervals.

portion of the

At the top of the graph is shown the observed integrated
energy release and the compilation of J. F. Perkins
and R. W. King (private communication) based on

spectral studies of chemically separated fission

products (use righthand scale).

s pectrometer

No correction for the

response function was

made for the
photon data and only a very approximate correction

was made for the energy release data.

to have been observed previously. Since the
energy release in a beta-ray decay would have
to be larger than typical neutron binding energies
if these gamma rays were to follow beta decay,
the short-period radiation described above is
assumed to arise from primary fission products
which are formed in isomeric states. While the
data given above only extend to 10~% sec, the
range to 10~3 sec has been surveyed. Presently
unavoidable random coincidence backgrounds were
larger than the counts which might have been
observed from any gamma radiation in this time
region.

These gamma rays follow fission by so short
an interval that they may be considered *'prompt”’

PERIOD ENDING MARCH 31, 1958

UNCLASSIFIED
7 2-0¢-058-0- 385

TIME AFTER FISSION (sec)

0425 x10 °

-sec”' Mev ™!

0.50 x 10

counts « figsion

 

o 0.2 o4 as 08 10 12 14 16 1.8 2.0 2.2
GAMMA - RAY ENERGY (Mev)

Fig. 4.3.3. Pulse-Height Spectra Observed at Short
Times After Fissions The top curve represents the

prompt radiations Random backgrounds have been

subtracted, and lines are fitted to the points only to

connect associated data.
for shielding purposes. However, experiments
planned to observe prompt radiation may exclude
these delayed radiations. As has been reported,”’
gamma rays emitted during this period may be
important in the consideration of the aftereffects
of nuclear explosions.'!

Prompt Gamma Rays

Following a period in which extensive checks
were made’ to ensure credible results, many
have been obtained for
of the spectrum of gamma rays

measurements a new

determination

 

's. Glasstone (ed.), The Effects of Nuclear Weapons,
p 343, USAEC, Washington, D. C., 1957.

107

 
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
2-01-058-0-386

‘T AYERAGE ENERGY ENERGY RANGE

(Mev) {Mev)

015 —0.22
0.25—0.35
0.48—092
1.48—1.42

0.19 ae
Q.20 a4
0.70 a
1.20 v

-
‘g8ec

—1

counts - fission

 

0 20 40 60 80 100 (x107%)

TiME {sec)

Fig- 403-4'
Gamma Rays Within Selected Energy Groups for Times
Less than 10~ % sec After Fission.

been made for random backgrounds and the prompt

Time Dependence of the Emission of
Corrections have

gamma-ray contributions. Data have been combined for
adjacent channels on the time analyzers Before combi-
nation, the time-channe! width for most of the runs was
(9.3 + 0.:2) x 1078 sec.

given energy group correspond to independent runs.

The different symbols for a

The errors shown are determined from the number of
recorded counts, and the lines are drawn only to connect

related points.

emanated in coincidence with the thermal-neutron
fission of U?33, Additional experimental results
will be required to allow a final evaluation of
these data, but a preliminary analysis was performed
in which compensation was made for the nonunique
response of the spectrometer to monoenergetic
gamma radiation. The results of this new experi-
ment, after the approximate analysis, are shown
in Fig. 4.3.5. The results are in fair agreement
with those obtained in a previously reported
preliminary expe.-rimenf,3 and in general agreement
with those of Gamble,'? although they differ as
much as a factor of 2 in some energy regions.

The data shown in Fig. 4.3.5 may be integrated
to yield 7.2 * 0.8 Mev/fission and 7.4 *+ 0.8
photons/fission

in the energy range between

 

12y E. Francis and R. L. Gamble, Phys. Semiann.
Prog. Rep. March 20, 1955, ORNL-1879, p 20.

108

Table 4.3.1. Fission Gamma-Ray Delay Fraction

as a Function of Energy*

 

 

Average Gamma-Ray
Gamma-Ray Energy ND/NP**
Energy Interval
(Mev) {Mev)
0.]9 0!15—'0-22 0-20 i 0004
0.30 0.25-0.35 0.07 +0.03
0.70 0.48--0.92 0.038 +0.01
1.3 1.18-1.42 0.020 *0.004
Integral 0.16—1.93 0.057 +0.003

 

*These data and the quoted errors apply to count-rate
data as obtained in pulse-height groups from a small
single crystal, Interpretation of these results as true
photon inténsities might lead to errors of a factor of 2 or
more.

**ND = total number of counts per fission observed in
the specified energy intervol for delay times between
5 % 10~ 8 and 10— sec; Np = number of counts per fis-
sion associated with prompt gamma rays in the same

energy interval,

0.3 and 10 Mev. The error estimates in Fig.
4.3.5 and in the integral quantities are preliminary,
since an accurate error analysis is impossible
until the experiment has been concluded.

It is expected that future experiments will
increase the accuracy of the data for energies
1.5 Mev.
be performed to yield data between about 80
and 800 kev. The results of a thorough analysis
of the existing and anticipated data will be made
available as soon as possible.

above An additional experiment may

TOTAL-ABSORPTION GAMMA-RAY
SPECTROSCOPY

G. T. Chapman T. A, Love

A series of gamma-ray spectral measurements
important for ANP shielding design is planned
for both reactors at the BSF, the aluminum-clad
BSR-l and the stainless-steel-clad BSR-Il. A
new spectrometer system, called the BSR model 1V

gamma-ray spectrometer,'3 which is now under

 

13R. W. Peelle, T. A. Love, and F. C. Maienschein,
ANP Quar. Prog. Rep. March 31, 1957, ORNL-2274
(Part 6), p 27.

 
20

-1
- Mev

—1

photons - fission

PERIOD ENDING MARCH 31, 1958
UNCLASSIFIED

2-01-058-0-388

¢ COMPTON SPECTROMETER

o PAIR SPECTROMETER

MEASURED ENERGY RESOLUTION (FULL PEAK WIDTH AT HALF HEIGHT}

i 2

PAIR SPECTROMETER — =

2.0 3.0

GAMMA ~

Fig- 4.3.5.
shown were obtained from counting statistics, and the
over which the results were averaged. The resoclution

horizontal barss The line is drawn only to connect the

 

Energy Spectrum of Gamma Rays Observed Within 10~7 sec After Fission.

—— COMPTON SPECTROMETER

T
Average photons - fission -Mev 1
from 7.7 to 10.5 Mev

4.0
RAY ENERGY (Mev)

5.0 6.0 7.0

The ordinate errors
energy errors represent in each case the energy interval
functions of the spectrometers used are indicated by the

pointss This plot represents a preliminary analysis of the

dota, and systemotic errors as large as 15% may occur in some ener regions.
’ Yy g g g

construction, will be used for these measurements,
and it is hoped that the detector for the system
will be a total
Such a crystal

interpretation of complex gamma-ray spectra owing
to the increased probability that a gamma ray
may lose nearly all of its energy in the crystal.
This, effect, would simplify the observed
pulse-height distribution for any given gamma-ray

absorption’’ Nal(TIl) crystal.
potentially may facilitate the

in

energy by eliminating a major portion of those
lower-energy pulses arising from scattered photons
escaping from the crystal.

Currently, multiple-crystal spectrom-
eters'4 are used for obtaining reactor spectral
data. Unfortunately, these spectrometers are

various

 

MF. C. Maienschein, Multiple-Crystal Gamma-Ray
Spectrometer, ORNL-1142 (July 3, 1952).

109

 
ANP PROJECT PROGRESS REPORT

complicated electronically and have low effi-
ciencies which require time-consuming operations.
The total-absorption spectrometer, with its greater
sensitivity, should simplify spectral measure-
ments through the possible elimination of the
complicated two-speed coincidence and pulse-
height analysis system required with the muitiple-
crystal spectrometers,

The total-absorption crystal used in the model 1V
spectrometer system will probably be a 9-in.-dia
right cylindrical crystal of Nal(Tl), one of which
is currently being investigated ot the BSF.'3
A %-in.-dia by 2-in.-deep well was drilled into
the truncated end of the crystal (Fig. 4.3.6) in
order to reduce the effects on the spectral
response of Compton scattering near the surface
where the gamma radiation enters the crystal.

The crystal has been investigated by introducing
gamma rays of energies ranging from 0.279 Mev

 

3g. T, Chapman, ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 295; G. T. Chapman and T. A.
Love, ANP Quar. Prog. Rep. March 31, 1957, ORNL.-
2274 {(Part 6), p 25.

   
  

MOUNTING RING

NEOPRENE REFLECTOR
SUPPORT

3/4-in.-DIA x 2-in-DEEP WELL

(ngoa) to 2.76 Mev (Na?4) into the truncated
end of the crystal through long lead collimators.
The response of the crystal to the two gamma
rays from Na?4 before the well was drilled into
the crystal is shown in Fig. 4.3.7. The pulse-
height distribution was obtained with a 20-channel
pulse-height analyzer. The distortion that appears
on the low-energy side of the 2.76-Mev peak was
noticeable when any gamma ray of energy greater
than about 1.6 Mev was introduced into the crystal.
It was postulated that this distortion was due
to the interaction of gamma rays near the surface
of the crystal and to subsequent escape of the
scattered gamma rays from the small-diameter
region of the cone.

The response of the crystal to the same gamma-
ray energies with the same conditions, but with
the gamma rays collimated into the bottom of the

well, is shown in Fig. 4.3.8. The shape of the

distribution curve indicates an improvement over
the distribution shown in Fig. 4.3.7 in that the
2.76-Mev

peak is more symmetrical about the

UNCLASSIFIED
2-01-058-0-203

_—FOR EQUIPMENT
ATTACHMENT

\\\\\\m

N

—
!

S
N

i

iy
%/\

.

l

{/

¢

Fig. 4.3.8. Sketch of 914-in.-dia Sodium lodide (Thallium-Activated) Gamma-Ray Spectrometer Crystal Showing

:z-in.-diu by 2=in«~Deep Well in Truncated End.

110

 
UNCLASSIFIED
2-01-058-0-164

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

300 I F
s 137 Mev
280 —— ——d o ——
260 U O I . v ]
240 : : e e
220 Ll . Lor i, C e e R
NOTE: DASHED CURVE
= 200 L J1! sHows PossiBLE CoMPTON |
E SCATTERING CONTRIBUTION.,
5
> 180 , : { L e
<
E
5 160 |- : e e
@
<
w140 bW 1 276 Mev
'_ ®
a {
o .
Q 420 | - : -
=
% »
3 100 [ - H-— - -4
Q &
80 ! RESOLUTION ™R T |
1% —of V| t- ]
60 AU f . ll,, A
7 ;oo
»
40 —— 7 i ,,,15,,,,,, ll‘ ‘.
7 i \
20 P k__ .‘{ o
LA . o \ \
I nt' \
0 b Vereee
0 10 20 30 40 50 60 70 80 90

RELATIVE PULSE HEIGHT

Fig. 4.3.7. Response of the 9/-|n.-d|o Crystal to
2.76- and 1.37-Mev Na24 Gamma Ruys Collimated into
the Truncoted End of the Crystal. The distortion
on the low-energy side of the 2.76-Mev peak was
postulated to be the result of multiple interactions in

the small-diameter region of the cone.

peak channel. A nearly constant resolution of
about 15% was observed for all energies during
these tests. This effect has not been explained
as yet.

In an attempt to improve the percentage resolu-
tion, seven 3-in.-dia photomultiplier tubes were
mounted on the crystal. A voltage divider was
constructed to allow an accurate adjustment of
the individual photomultiplier tube gains and
data were obtained with the 256-channel analyzer.
The response to the Na?4 gamma rays collimated
into the well, as before, is shown in Fig. 4.3.9.
The shape of the curve indicates an improvement

PERIOD ENDING MARCH 31, 1958

UNCL ASSIFIED
ORNL—LR—DWG 23045

]

i

70 ; ‘
6o L 1 J‘4.37 Mev ‘
50 ’/ ]

40 1 S

 

 

 

 

 

 

30 —- ‘L, - g —

oo

10
s * =g J
a A g

 

 

 

 

COUNTING RATE (ARBITRARY UNITS)

 

 

 

 

 

 

!
L

 

 

 

 

 

 

 

oo T ke
e
0 20 40 60
PULSE HEIGHT
Fig. 4.3.8. Response of the 9/-tn.-d|u Sodium

lodide Crystal to Na?4 Gamma Rays Callimated into
the Bottom of the Well and Observed with Three 5-in.-dia
Photomultiplier Tubes. The ratio of the peak area to
the tatal area is 0.73 for the 1.37-Mev gamma ray and

0.67 for the 2.76-Mev gamma ray.

in that the distribution is not so flat near the
peak channels. It is believed that the addition
of the well improved the response. (The new
256-channel  pulse-height analyzer has only
recently been put into operation, It is necessary,
because of the variable dead time of the analyzer,
to install an automatic dead-time clock to
facilitate the accurate subtraction of the back-
ground from the data. This equipment is being
installed in the analyzer.)

A comparison of the photofractions, that is, the
ratio of the area under the full-energy peak to
the area under the total distribution curve, ob-
tained with the well in the crystal with those
calculated by Berger and Doggett'® is presented
in Fig. 4.3.10. The calculations were made by
using a Monte Carlo technique for a 5-in.-dia
by 9-in.-long cylindrical crystal. The measured
photofractions agree to within about 3.5% with
the calculated values. Since in both cases the

gamma rays were in the energy region below

 

]6M J. Berger and J. Doggett, |. Research Natl. Bur.
Standards 56, 355 (1956).

11

 
ANP PROJECT PROGRESS REPORT

3 Mev, and since the radial loss is low for crystals
as large as 5 in. in diameter, it is believed that
the comparison is probably realistic, regardless
of the difference in crystal geometry. A com-
parison with other existing spectrometers is pre-

 

 

 

 

 

 

 

sented in Table 4.3.2. This table shows the

UNCLASSIFIED
2-01-058-0-387

4000 1.37 Mev

3600

3200 3

2800 4

2400

P 4
| 2.76 Mev
2000 - o

 

 

 

COUNTING RATE (arbitrary units)

1600 ——- X
1200 ® .
| \
4
800 -

 

 

 

 

 

 

 

 

 

 

 

 

 

 

%
@
'0'\'\'1& .
400
ag?
k‘ofi‘l’:‘. i k
0 %,
0 20 40 60 80 100 120 140

RELATIVE PULSE HEIGHT

Fig. 4.3.9. Response of the 9/ in.~-dia Sodium lodide
to Na24 Gamma Ruys Collimated
Bottom of the Well and Observed with Seven 3«in.-dia
Photomultiplier Tubes.

Crystal into the
Mo measurement of the photo-
fraction was attempted since the equipment did not

allow an accurate background subtraction.

measured photofractions for the total-absorption
spectrometer and those measured with multicrystal
spectrometers. '4 The values for the multicrystal
spectrometers were obtained from a rough fit to
data taken by Peelle, Love, and Zobel.17

The total-absorption spectrometer, with the
present crystal, does not compete favorably with
the other spectrometers in terms of resolution,
The efficiency of the total-absorption spectrometer
is, however, two to three orders of magnitude
greater than the efficiencies of the multiple-

crystal spectrometers,

 

7W. Zobel, private communication.

UNCLASSIFIED
ORNL-LR-DWG 23046

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.16
CZ) /D
=
3 045 -
o
)
wl
: |

044

0.9 \

T EASURED WITH 9 ¥g-in.~dio CRYSTAL

0.8 |- !
~ do
'Lu —
w 7 e

' T
CALCULATED FOR S5-in,—dig x 9—in.
0.6 CRYSTAL, NBS 56,335 (1956) —
0.5 ‘
O 0.4 0.8 1.2 1.6 2.0 2.9 2.8
£, (Mev)

Fig. 4.3.10. Ratio of Peak Area to Total Areq, p(E )s
for the 9/-|n.-d|c| Sodium lodide Crystal as a Funchon
of the Energy of Gamma Rays Collimated into the Well.

Table 4.3.2, Photofractions Obtained with Several Spectrometers for Gamma Rays of Various Energies

 

Photofractions

 

 

Gamma-Ray

Energy Total-Absorption Pair Compton
(Mev) Spectrometer Spectrometer Spectrometer
0.662 0.87 0.81
0.903 0.74 0.76
1.37 0.73 0.99 0.67
1.85 0.69 0.94 0.57
2.76 0.67 0.84 0.38

 

112

 
A STUDY OF RADIATION HEATING IN SHIELDS
K. M. Henry

experimental work with high-efficiency
has been devoted almost entirely to

Past
shields
measurements of the radiation attenuation of the
shields, with little investigation of the con-
comitant heating produced in the shields by the
radiation. Since it may be necessary to provide
cooling channels in aircraft shields, a study of
the heat production in a large slab mockup of a
typical shield was undertaken at the BSF. This
work is now in progress, with the objectives of
the experiment being (1) to obtain experimental
information that can be used to verify the results
of the best available methods for calculating
the energy deposition in a shield, (2) to obtain
data that can be used directly in current shielding
problems, and (3) to obtain flux and dose measure-
ments within a shield so that correlations can
be made between these measurements and heat
deposition.

The shield mockup currently being investigated

2-01-058-0-391

 

 

OIL TANK; 3/g-in.-THICK
ALUMINUM WALLS

SLABS 4 ft SQUARE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

WATER
- s % 5 ft SQUARE
3‘/2in. I g
=|a
2@
=T I
&5 3
1k
3% - INSTRUMENT
K{ 67 HOLES 1, 2, & 3
IS BT Tl waTer
Ffi§-—HEATING
SAMPLES
BULK L
SHIELDING
REACTOR c £ -
(LOADING NO. 33) {| [T [ <] & -
sin.—a] b=
GULF 10-C
INSULATING OIL
GAMMA SHIELDING MATERIALS
% in. OF IRON
. ) —_— A —
3in. OF LEAD

2 in. OF MALLORY 1000

Plon View of BSF Shield Heating

Figo 4.3.]].

Experiment with Heating Samples in the Fore Position.

PERIOD ENDING MARCH 31, 1958

consists of a 4-in.-thick slab of beryllium, fol-
lowed by a gamma-ray shield and 16 in. of lithium
hydride. The lithium hydride is divided into two
sections of 4 and 12 in., respectively, as shown
in Fig. 4.3.11.
varies in composition, consists of 3 in. of lead,
3 in. of iron, or 2 in. of Mallory 1000. The entire
shield is submerged in an oil-filled aluminum
tank positioned against the Bulk Shielding Re-
actor, with the horizontal axes of the shield and
reactor coincident.

Measurements of the heating within the shield
caused by radiation from the reactor are obtained
by attaching thermocouples on or near thin disk-
type samples located within the shield. The heat
loss, or cooling of the shield, is also observed
after the reactor is removed from the vicinity of
the shield.

The heating-measurement samples, which are

The gamma-ray section, which

each 1.25 in. in diameter, are encased in con-

tainers made of the same material (see Fig.

4.3.12). The samples themselves are surrounded
e i

UNCLASSIFIED
2—01—-058—0- 392

0.037-in.-0D CERAMIC
SUPPORT TUBES, 0.009-in.
THICK WALLS B

      

1.25-in,
A DIAMETER
EXTERNAL VIEW SECTION A-A
FROM EDGE
- 2in.

i - CERAMIC FEED-THROUGH INSULATOR
1M FOR THERMOCOQUPLE LEADS {(SEALED
IN WITH "DEVCON B")

 

 

 

 

 

 

 

 

 

i ® LOCATION OF THERMOCOUPLES
- SAMPLES
MATERIAL  THICKNESS
r (@ IRON 0.2 in.
1 e LEAD 0.4 in.
D & £l £l MALLORY 1000 0.1 in.
/ / ~NF 712 BERYLLIUM 0.4 in.
O O

 

 

 

 

% vrw,,L

TWO HALVES OF CONTAINER CEMENTED
TOGETHER WITH "DEVCON B PLASTIC STEEL"

 

SECTION B-B

Details of Metallic Heating Somples,
Locations for BSF

Figo 4-3.]2.
Containers, and Thermocouple

Heating Experiment.

113

 

 
ANP PROJECT PROGRESS REPORT

by layers of air except at points where support
tubes are attached. (Because of fabrication
difficulties, it is necessary to encase the lithium
hydride sample in a cadium-covered aluminum
case.) Each sample case contains four thermo-
couples, one attached to each face of the sample
and one attached to each inside face of the sample
case. The cases are plugged into recessions in
the faces of the shield sections along the shield

axis. In a specific shield assembly all the
samples are either on the sides of the shield
sections facing the reactor (fore position) or

on the sides away from the reactor (aft position).
The diameters of the shield cases used in the
aft position are larger than the diameters of
those used in the fore position so that radiation
streaming through crevices straight through the
shield can be avoided. At the interfaces of the
shield sections, vertical wells extend downward
through the shield to the samples. Flux and dose

AR
6 2-01-058-0-393

 

7= TIME, p=POWER (ARBITRARY)
DETECTOR: U235 FiSSION CHAMBER

HOLE { {Be—Fe INTERFACE)

HOLE 2 (Fe—LiH INTERFACE)

THERMAL—NEUTRON FLUX (neutrons-cm =2+ 1. p71)

HOLE 3 {LiH—-LIH INTERFACE)

o 4 8 12 16 20 24
DISTANCE ABOVE SHIELD AXIS (in)

Figl 4-3-]3-
Interfaces os a Function of Distance Above the Shield

Axis (BSF Shield Heating Experiment).

Thermol-Nevtron Fluxes at the Shield

114

rate measurements are made by instruments
positioned within these wells (U?3% fission
detector, Hurst dosimeter,and ionization chambers).

The wells also accommodate the thermocouple
leads.

All planned measurements with configurations in
which lead and iron were used as the gamma-ray
shield have been completed. In addition, measure-
ments with configurations using Mallory 1000 have
been completed for the samples in the fore
position. The work has been delayed temporarily
in order to permit the Mallory 1000 activity to
decay sufficiently so that the shield can be
approached.

The thermal<neutron fluxes, the fast-neutron
dose rates, and the gamma-ray dose rates at
the beryllium-iron, iron-LiH, and LiH-LiH inter-
faces in the beryllium-iron-LiH configuration are
given in Figs. 4.3.13, 4.3.14, and 4.3.15, re-
spectively., Measurements of the temperature in
the beryllium, iron, and lithium hydride samples
are shgwn in Figs. 4.3.16, 4.3.17, and 4.3.18,
respectively. These data are presented only to
indicate orders of magnitude; no attempts have
yet been made to analyze flux, dose, and tempera-
ture data in terms of actual reactor power.

2—01—055-0—394

103
/=TIME; p=FPOWER (ARBITRARY)
5 |- DETECTOR: HURST DOSIMETER

T
Q2
v 402
w HOLE f (Be—Fe INTERFACE)
o
.o 5
1 | ol
= HOLE 2 (Fe—LiH INTERFACE)
L2
o
L0
w
b
2 s
w
u
2
> HOLE 3 (LiH—LiH INTERFACE)
g
._
o
z 5
l._
w
o 2

 

0 4 8 12 16 20 24
DISTANCE ABOVE SHIELD AXIS (in}

Fig. 4.3.14. Fost-Neutron Dose Rates ot the Shield
Interfaces as a Function of Distance Above the Shield
Axis (BSF Shield Heating Experiment).

 

 
SGRapHITE * /

-1

GAMMA-RAY DOSE RATE (ergs -

Fig- 4-3-]5.

 

PERIOD ENDING MARCH 31, 1958

A
2-01-058-0-395

f=7TIME; p=POWER (ARBITRARY}
DETECTOR: IONIZATION CHAMBER

HOLE { (Be—Fe INTERFACE)

HOLE 2 (Fe—LiH INTERFACE)

HOLE 3 (LiH-LiH INTERFACE )

4 8 12 16
DISTANCE ABOVE SHIELD AXIS (in)

20 24

Gamma-Ray Dose Rates at the Shield

Interfaces as o Function of Distance Above the Shield

Axis (BSF Shield Heating Experiment).

2-01-058-0-396

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

* | |
l-—— REACTOR AT CONSTANT POWER ——-——J r/REACTOR SCRAMMED
{HEAT SOURCE REMOVED)
30 || |
32 | f
30 { A A ,...
o f |\\y—-8e SAMPLE TEMPERATURE (TWO RUNS)
o
& Be CASE TEMPERATURE (TWC RUNS)
20 || 7 %
: R |
g | /
* | )/U |
26 ! ,
P |
| .7 |
24 lfi {
22 g :
L |
0 1:00 2:00 300 4:00 5:00 6:00 7:00
TIME (hr}

Figu 4-30]6.
Shield Heating Experiment).

Temperature of the Beryllium Sample at the Be«Tank Wall Interface as a Function of Time (BSF

115

 
ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

L 2-01-058-0-397
36
REACTOR AT CONSTANT POWER — REACTOR SCRAMMED
(HEAT SOURCE REMOVED)
34 '
’ A ‘
| 3}
| \\/‘/—Fe SAMPLE TElMPERATURE (TWOiRUNS)
%0 ] Ty
5 [ ‘ 3 Fe CASE TEMPERATURE (TWO RUNS)
w
5 | \
E 28 o
< [
3 %
s | ,
P /4
[ | \ |
26 | 7, | | \q\c\
/ ! ER\D::
"1 | T
22 ; —~ [
20 — " ‘
0 e ® 00 2:00 3:00 4:00 5:00 6:00 7:00

TIME {hr)

Figs 4.3.17. Temperature of the Iron Sample at the Be-Fe Interface as a Function of Time (BSF Shield Heating
Experiment).

2-01 -058—!—398

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

30 | |
| REACTOR SCRAMMED
28 (HEAT SOURCE REMOVED)
|——wm REACTOR AT CONSTANT POWER *—_—l ‘
0 l /LiH SPECIMEN TEMPERATURE (TWO RUNS)
L 26 b
; =
D P
: |
@x
Ll
e
Z 24 |
o
I Cd —COVERED Al CASE TEMPERATURE
I l { TWO RUNS)
i
ol |
0 1:00 2:00 3:00 4.00 5:00 6:00 7:00

TIME (hr)

Figs 4.3.18. Temperature of the Lithium Hydride Sample at the Fe-LiH Interface as a Function of Time (BSF
Shield Heating Experiment).

116

 
PERIOD ENDING MARCH 31, 1958

4.4. TOWER SHIELDING FACILITY

L

C. E. Clifford

Applied Nuclear Physics Division

AIRCRAFT SHIELD TEST REACTOR
EXPERIMENT AT THE TOWER
SHIELDING FACILITY

Y. R. Cain J. L. Hull
B. J. Workman' F. J. Muckenthaler

In cooperation with Convair, Fort Worth, an
experiment is now being conducted at the Tower
Shielding Facility in which neutron and gamma-
ray dose rates and thermal-neutron fluxes are
being obtained under conditions similar to those
encountered with the Nuclear Test Airplane (NTA)
program but in the absence of the airplane structure.
The NTA program at Convair included shielding
measurements taken inside and outside the NTA,

 

]Convuir, Ft. Worth.

both while in flight and while on the ground. It
also included measurements in the absence of
the airplane with the Aircraft Shield Test Re-
actor (ASTR) on a cradle 12.5 ft above a concrete
pad. All these measurements will be compared
with measurements taken in air at the TSF to
determine magnitudes of ground and aircraft-
structure effects. |t will also be possible to
obtain a more accurate source term for the ASTR
in this experiment, since ground effects can be
greatly reduced.

The ASTR consists of an array of MTR-type fuel
elements mounted horizontally between two grid
plates. Demineralized water is used as moderator,
reflector, and coolant. The fuel element arrange-
ment can be seen in Fig. 4.4.1, which also shows

UNCLASSIFIED
ORNL -LR-DWG 28942

 

. GAMMA MONITOR

. AFT GRID

. PRESSURE VESSEL LEAD CTASE
. CONTROL ROD

. MAIN SHIELD TANKS

U W ) -

FUEL ELEMENTS
FORWARD GRID

LEAD RINGS AND DISKS
FORWARD SHIELD TANKS
ION CHAMBER

Som~e

Fig. 4.4.1. Cutaway View of the Aircraft Shield Test Reactor (ASTR).

117

 
ANP PROJECT PROGRESS REPORT

the location of the various reactor shield tanks,
which can be filled or drained remotely, and Fig.
4.4.2 shows the locations of the ASTR, the NTA
crew compartment (also shown in Fig. 4.4.3),
the cylindrical crew-shield mockup (also shown
in Fig. 4.4,4), and various detectors in the
positions used in the NTA,

The TSF experiments are being conducted in
three phases: (1) data are being obtained for
comparison with data taken at Fort Worth with
the reactor center line 12.5 ft above ground level;
(2) the spectra of neutrons from the ASTR are
being determined, and (3) measurements are being
made for comparison with NTA detector station
measurements.

The configuration used in Phase 1 is shown in
Figs. 4.4.5 and 4.4.6. The reactor support, which
is attached to the hoist cables, allows the re-
actor to rotate 180 deg. The support truss for
the detector and the crew compartment or crew-
shield mockup is suspended from two hoist cables
at one end and is attached to the reactor support
at the opposite end. In Phase 1, groups of one
dosimeter, one anthracene scintil-
lation counter, one bare BF3 counter and one
cadmium-covered BF3 counter are suspended
from the truss on the reactor center line at dis-
tances of 33, 59, and 100 ft from the reactor
center. The 100-ft detector station is shown in
Fig. 4.4.7. Measurements are made with the
various detectors as a function of reactor angular
orientation, with the center line of the system
12,5 ft above the concrete apron, as a function
of altitude, and as a function of reactor angular
orientation at full altitude.

fast-neutron

A neutron collimator assembly will be suspended

from the truss in Phase 2. Neutron emulsions

118

ORNL-LR-DWG 28943

 

 

 

PLAN VIEW

 

® GAMMA AND NEUTRON DETECTORS, DIRECT BEAM SHIELDS
4 GAMMA AND NEUTRON DETECTORS, NO DIRECT BEAM SHIELDS

   
  
  
 

CYLINGRICAL CREW-SHIELD MOCKUP

  
   

SHADOW SHIELD
ASTR

— e ——a—

INSTRUMENT CAPSULE
CREW COMPARTMENT

  

 

SIDE ELEVATION

NOTE: DATA TAKEN AT SiIX POSITIONS INSIDE THE
CREW COMPARTMENT

Fige 4.4s2. Detector, Crew Compartment, and Crew-
Shield Mockup Positions in the NTA.

will be placed within the collimator to determine
the fast-neutron spectra emitted from the ASTR
with the side tanks emptied.

In Phase 3, the NTA crew compartment and the
cylindrical crew-shield mockup will be suspended
from the truss at positions duplicating those found
in the NTA. Measurements with all types of de-
tectors will be made at all the positions used
in the NTA, as shown in Fig. 4.4.2, except those
in the tail section of the airplane.

 
PERIOD ENDING MARCH 31, 1958

 

UNCL ASSIFIED
PHOTO 43314

Figs 444.3. The NTA Crew Compartment.

 

119

 
ANP PROJECT PROGRESS REPORT

UNCL ASSIFIED
PHOTO 43315

"

120

 

4.4. The Cylindrical Crew-Shield Mockup Used in the NTA,

4.

Figl
UNCL ASSIFIED
PHOTO 43318

BEER L CEr E bl ol

 

PERIOD ENDING MARCH 31, 1958

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ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
PHOTO 43146

    

 

T TS T\ e:.-_ ha ..-gr I TIR—NNTR
7 4“ _fi'?i'lu s .I'-___.* Y ’lh ”.“nli H; g”“.';'

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e

122

Figs 4¢4.6. Closeup of the ASTR ond Two Groups of Detectors Suspended from the Truss.

 
PERIOD ENDING MARCH 31, 1958

N\ uncLassiFIED)
M= PHOTO 43317 |

 

Fig. 4.4.7. Detector Station 100 ft from the ASTR.

123

 
OVOONOOGE WN —
& e & & & e & & e 4w

NN NN MNMNAN — — ed od od o md =d =md
NN B W = O VO ON B WA —
. e . . . . . « e . . . . . .

27.
28.
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31.
32.
33.
34.
35.
36.
37.
38.
39.
40.
41,
42,
43.
44-45,

70&—&—:‘1106‘)5>>Im§—lnm>smx’s&—:br‘or';—o'nig—n;on>-n:orno£nm-nos—

W, Allen

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H. Coobs

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ORNL-2517
C-85 — Reactors-Aircraft Nuclear
Propulsion Systems

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orto
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O-sr N <comONO Z

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Erm-ppITC

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White

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W Zobel

ORNL - Y-12 Technical Library,
Document Reference Section

Laboratory Records Department

Laboratory Records, ORNL R.C.
Central Research Library

DEEmME-PEOEmME-Omp

125

 
EXTERNAL DISTRIBUTION

 

101-103. Nyr Force Ballistic Missile Divigin s
104-105. ANPR, Boeing, Seattle

106. AFRR, Boeing, Wichita

107. AFPN, Curtiss-Wright, Clifton *

108. AFPR\Douglas, Long Beach §
109-111. AFPR, Dguglas, Santa Monicg

112. AFPR, Lo\kheed, Burbank
113-114. AFPR, Lockgeed, Marietta

115. AFPR, North\merican, Ca

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132-137. Atomic Energy Comniil#sion, Wishington

138. Bettis Plant (WAPD

139. Bureau of Aeronauti

  
  
  
  
    

ha Park

 
  

 
    
   
 
 
   
 
  
  
  
  
 
  
    
  

vair, Fort Worth

  
   
 
 
 
 
  
 
  
 
   

126

 

140. Bureav of Aeronautfill General Rggpresentative
141. BAR, Aerojet-Genejlilk, Azusa
142. BAR, Convair, Sanliego
143. BAR, Goodyear Aiglaft, Akron
144, BAR, Grumman Aigllaft, Bethpage
145. BAR, Martin, Baltfllore
146. Bureau of Ships
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161. Marquardt Aircraftffll_ompany
162. National Advisoryl_ommittee for Aeronautics tCleveland ,
163. National Advisory ommittee for Aeronautics, ashington
164. Naval Air Develop ent Center
165. Naval Air Material Center .
166. Naval Air Turbine Test Station

 

 

 

 
167.
168.
169.
170.
171.
172-175.
176.
177.
178.
179.
180.
181.
182.
183.
184-185.
186-203.
204-218.
219.

  

fon, Livermore
bn Medicine

 
 
   
 
  
 

rnia Radiation Laboratory, Livermore
ent Center

Service Extension, Oak Ridge

and Development,

hission, Oak Ridge Operations

 

127
128

ORNL-528
ORNL-629
ORNL-768
ORNL-858
ORNL-919
ANP-60
ANP-65
ORNL-1154
ORNL-1170
ORNL-1227
ORNL-1294
ORNL-1375
ORNL-1439
ORNL-1515
ORNL-1556
ORNL-1609
ORNL-1649
ORNL-1692
ORNL-1729
ORNL-1771
ORNL-1816
ORNL-1864
ORNL-1896
ORNL-1947
ORNL-2012
ORNL-2061
ORNL-2106
ORNL-2157
ORNL-2221
ORNL-2274
ORNL-2340
ORNL-2387
ORNL-2440

.

Reports previously issued in this series are as follows:

Period Ending November 30, 1949
Period Ending February 28, 1950
Period Ending May 31, 1950
Period Ending August 31, 1950
Period Ending December 10, 1950
Period Ending March 10, 1951
Period Ending June 10, 1951
Period Ending September 10, 1951
Period Ending December 10, 1951
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Period Ending September 10, 1952
Period Ending December 10, 1952
Period Ending March 10, 1953
Period Ending June 10, 1953
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Period Ending December 10, 1953
Period Ending March 10, 1954
Period Ending June 10, 1954
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Period Ending March 10, 1955
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Period Ending December 31, 1956
Period Ending March 31, 1957
Period Ending June 30, 1957
Period Ending September 30, 1957
Period Ending December 31, 1957